ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 59

DOE Review Release 07/27/2011

Document Title:

Pinellas Plant – Occupational External Dose

Document Number: ORAUT-TKBS-0029-6

Revision: 01

Effective Date: 04/28/2011

Type of Document: TBD

Supersedes: Revision 00 PC-2

Subject Expert(s): Brian P. Gleckler and Peter A. Darnell (NIOSH)

Approval: Signature on File Approval Date: 04/27/2011

Brian P. Gleckler, Document Owner

Concurrence: Signature on File Concurrence Date: 04/26/2011

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 04/26/2011

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 04/26/2011

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 04/28/2011

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 2 of 59

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

09/15/2005 00 New Technical Basis Document for Pinellas – Site External Dose. 

First approved issue. Training required: As determined by the Task 

Manager. Initiated by Mark D. Notich.

08/03/2006 00 PC-1 Approved page change revision to incorporate changes according to 

the public outreach meetings in 09/2004 and 11/2005 and claimant 

communications. Updated required language on pages 6 and 7 in 

Section 6.1. Added Purpose and Scope Sections on page 7. 

Changes occurred on page 10, in Section 6.1.4, page 27 in Section 

6.4.2, and on pages 29 - 32 in the Reference Section. As a result of 

internal formal review, changes were made to Tables 6-4, 6-5, 6-11, 

and 6-12 on pages 16, 17, 25, and 27, respectively in Sections 6.2.3 

and 6.4.1.1 and 6.4.2. Wording was added or changed on pages 10, 

15, 19, 23, 24, and 27, in Sections 6.1.1, 6.2.3, 6.2.4, 6.3, and 

6.4.2.1. No sections were deleted. This revision results in a 

reduction in assigned dose and no PER is required. Training 

required: As determined by the Task Manager. Initiated by Paul J. 

Demopoulos. Approval:

Signature on File 07/25/2006

Paul J. Demopoulos, TBD Team Leader

Signature on File 07/25/2006

John M. Byrne, Task 3 Manager

Signature on File 07/24/2006

Edward F. Maher, Task 5 Manager

Signature on File 07/27/2006

Kate Kimpan, Projector Director

Signature on File 08/03/2006

James W. Neton, Associate Director for Science

11/08/2006 00 PC-2 Approved page change revision to incorporate changes according to 

Task 5 comment and Task 7 amendment to page 27 and 28 in 

Section 6.4.1.1. Corrects table numbering on page 28 and 29. Adds 

two references to page 34. Corrects GEND references on pages 20, 

24, and 32. No sections were deleted. Incorporates internal formal 

review comments. Incorporates NIOSH formal review comments. 

Minor editorial changes were made on pages 24 and 29. Deletes 

Task 7 amendment to page 27 and 28 in Section 6.4.1.1. Corrects 

Table 6-5 rows 2 and 3 on page 18. This revision results in an 

increase in assigned dose and a PER is required. Training required: 

As determined by the Task Manager. Initiated by Paul J. 

Demopoulos. Approval:

Signature on File 10/19/2006

Paul J. Demopoulos, Document Owner

Signature on File 10/20/2006

John M. Byrne, Task 3 Manager

Signature on File 10/24/2006

Edward F. Maher, Task 5 Manager
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 3 of 59

Signature on File 10/23/2006

Kate Kimpan, Projector Director

Brant A. Ulsh Signature on File for 11/08/2006

James W. Neton, Associate Director for Science

04/28/2011 01 This technical basis document was predominantly revised to address 

SC&A’s issues with the document, as identified in SCA-TR-TASK1-

0015. Adjustments to certain photon dosimeter results now need to 

be made per Section 6.3.1.1. Photon LOD values specific to the 

Mound dosimeters are now provided in Table 6-10. Significant 

changes were also made on how neutron doses are assessed, and 

the new approaches are described in Section 6.3.2. The applicability 

dates for the various types of neutron dosimeters used at the site 

were changed, and are now in Table 6-11. In addition, a number of 

editorial changes were made to this document. Incorporates formal 

internal and NIOSH review comments. Constitutes a total rewrite of 

the document. Training required: As determined by the Objective 

Manager. Initiated by Brian P. Gleckler.
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 4 of 59

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ...............................................................................................................6

6.0 Occupational External Dose.....................................................................................................8

6.1 Introduction ..............................................................................................................................8

6.1.1 Overview ......................................................................................................................9

6.1.2 Purpose........................................................................................................................9

6.1.3 Scope...........................................................................................................................9

6.1.4 Sources of Radiation .................................................................................................. 10

6.1.5 Occupational Dosimetry Program Overview ............................................................... 11

6.2 Dose Reconstruction Parameters...........................................................................................13

6.2.1 Historical Administrative Practices.............................................................................. 13

6.2.1.1 Performance Testing................................................................................. 17

6.2.1.2 Reports ..................................................................................................... 17

6.2.2 Dosimetry Technology................................................................................................ 18

6.2.3 Dosimetry Calibration Practices.................................................................................. 20

6.2.4 Workplace Radiation Fields ........................................................................................ 21

6.2.4.1 Electron and Photon Exposures ................................................................ 24

6.2.4.2 Neutron Exposures ................................................................................... 24

6.3 Adjustments to Recorded Dose..............................................................................................26

6.3.1 Photon Dose Adjustments .......................................................................................... 27

6.3.1.1 Signal Fading Adjustments........................................................................ 27

6.3.2 Neutron Dose Adjustments......................................................................................... 27

6.3.2.1 Radiation Weighting Factor Adjustments................................................... 27

6.3.2.2 Track Fading Adjustments......................................................................... 28

6.3.2.3 Signal Fading Adjustments........................................................................ 29

6.3.2.4 Neutron Energy Response Adjustments.................................................... 29

6.3.3 Electron Dose Adjustments ........................................................................................ 30

6.4 Missed Dose ..........................................................................................................................30

6.5 Unmonitored Dose .................................................................................................................33

6.5.1 Unmonitored Workers................................................................................................. 33

6.5.2 Workers Whose Dosimetry Records Are Missing........................................................ 34

6.6 Uncertainty.............................................................................................................................34

6.6.1 Beta-Gamma Dosimeter Uncertainty .......................................................................... 35

6.6.2 Neutron Dosimeter Uncertainty................................................................................... 35

6.7 Attributions and Annotations ..................................................................................................35

References ........................................................................................................................................36

Glossary ............................................................................................................................................43

ATTACHMENT A, EXAMPLES OF PINELLAS PLANT EXPOSURE RECORDS .............................. 49

ATTACHMENT B, BASIS FOR THE UNMONITORED EXTERNAL DOSE ASSIGNMENT ............... 54
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 5 of 59

LIST OF TABLES

TABLE TITLE PAGE

6-1 Personnel radiation dosimetry from AEC annual reports ........................................................ 12

6-2 Pinellas Plant historical dosimetry events............................................................................... 14

6-3 Dosimetry used at Pinellas Plant for external whole-body and extremity 

exposures ..............................................................................................................................18

6-4 Radionuclides contributing to workplace radiation fields......................................................... 23

6-5 Electron and photon radiation energies and percentages for Pinellas Plant 

processes ..............................................................................................................................24

6-6 Expected neutron radiation energies for Pinellas Plant processes ......................................... 25

6-7 Neutron radiation energies and ICRP 60 correction factors.................................................... 28

6-8 Summary of neutron dosimeter correction factors .................................................................. 30

6-9 Whole-body beta-gamma dosimetry missed doses ................................................................ 31

6-10 Whole-body neutron dosimetry missed doses........................................................................ 32

6-11 Extremity beta-gamma dosimetry missed doses .................................................................... 33

B-1 Evaluation of Pinellas Plant doses available in the SRDB ...................................................... 55

B-2 Evaluation of POCs associated with assigning unmonitored doses as 

either photon or neutron dose ................................................................................................ 59

LIST OF FIGURES

FIGURE TITLE PAGE

6-1 Regenerated example of a Landauer dosimetry report from the 1970s .................................. 16

6-2 RTG plutonium source neutron energy spectrum – PuO2 microspheres................................. 26
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 6 of 59

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

ALARA as low as reasonably achievable

Bq becquerel

CFR Code of Federal Regulations

Ci curie

cm centimeter

CPE Ceramic Product Engineering

D deuterium (

2

H)

DCF dose conversion factor 

DOE U.S. Department of Energy

DOELAP U.S. Department of Energy Laboratory Accreditation Program

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ft foot

g gram

GE General Electric Company

GEND GE Neutron Devices 

GENDD GE Neutron Devices Department 

GEPP GE Pinellas Plant

GEXF GE X-ray Division in Florida

GEXM GE X-ray Division in Milwaukee, Wisconsin

hr hour

HSR Health and Safety Record (system)

ICRP International Commission on Radiological Protection 

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

Landauer R. S. Landauer Jr. & Co.

LOD limit of detection

MeV megaelectron-volt, 1 million electron-volts

mR milliroentgen

mrem millirem

n neutron

NIOSH National Institute for Occupational Safety and Health

NOCTS NIO Office of Compensation Analysis and Support (OCAS) Claims Tracking System

NDD Neutron Devices Department

NTA nuclear emulsion type A film (a film used for certain neutron dosimeters)

ORAUT Oak Ridge Associated Universities Team
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PAO Pinellas Area Office

POC probability of causation

QA quality assurance

RBE relative biological effectiveness

RTG radioisotopically-powered thermoelectric generator

SNL Sandia National Laboratories

SRDB ORAUT’s Site Research Database

SRDB Ref ID Site Research Database Reference Identification (number)

T tritium (

3

H)

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

yr year

μCi microcurie

α alpha particle

§ section or sections
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 8 of 59

6.0 OCCUPATIONAL EXTERNAL DOSE

6.1 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation0F

1

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation, 42 C.F.R. Pt. 82) define “performance of duty” for 

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2010).

The statute also includes a definition of a DOE facility that excludes “buildings, structures, premises, 

grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 U.S.C. 

7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. While this 

definition excludes Naval Nuclear Propulsion Facilities from being covered under the Act, the section 

of EEOICPA that deals with the compensation decision for covered employees with cancer [i.e., 42 

U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally-derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external occupational radiation exposures are considered valid for inclusion in a dose reconstruction. 

No efforts are made to determine the eligibility of any fraction of total measured exposure for inclusion 

in dose reconstruction. NIOSH, however, does not consider the following exposures to be 

occupationally derived (NIOSH 2010):

• Background radiation, including radiation from naturally occurring radon present in 

conventional structures

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 9 of 59

6.1.1 Overview

This TBD is Part 6 of the Pinellas Plant’s Site Profile. A site profile provides a summary of information 

about a site that is relevant to the dose reconstruction process. 

The Pinellas Plant has been known by several names throughout its history. Those names include: 

908 Plant, Pinellas Peninsula Plant, GE X-ray Division-Florida (GEXF), GE Neutron Devices 

Department (GENDD), GE Neutron Devices (GEND), GE Pinellas Plant (GEPP), and the Pinellas 

Plant. 

The General Electric Company built and operated the Pinellas Plant for DOE from its initial startup in 

January 1957 until June 1992. In June 1992, Martin Marietta Specialty Components, Inc. (MMSC) 

took over as the managing and operating contractor for the Pinellas Plant. In 1994, Lockheed merged 

with Martin Marietta and the managing and operating contractor for the Pinellas Plant was renamed 

Lockheed Martin Specialty Components (LMSC). The Pinellas Plant completed its war reserve 

fabrication of neutron generators at the end of September 1994, and began the transition from a 

defense mission to an environmental management mission. That transition included a number of 

decontamination and decommissioning activities that allowed the Plant to be turned over for 

commercial uses. LMSC continued as the managing and operating contractor until decontamination 

and decommissioning activities ended in 1997 (ORAUT 2011b).

The Plant was built to manufacture neutron generators, a principal component in nuclear weapons. 

The neutron generators consisted of a miniaturized linear ion accelerator assembled with pulsed 

electric power supplies. The ion accelerator, or neutron tube, required ultraclean, high-vacuum 

technology; hermetic seals between glass, ceramic, glass-ceramic, and metal materials; and highvoltage generation and measurement technology. The Plant manufactured only neutron generators 

for its first 10 years of operation. It later manufactured other products including neutron detectors, 

radioisotopically-powered thermoelectric generators (RTGs), high-vacuum switch tubes, specialty 

capacitors, and specialty batteries (Weaver 1990). As part of its program to promote commercial 

uses of the site, DOE sold most of the Plant to the Pinellas County Industry Council in March 1995 

and leased back a portion through September 1997 to complete safe shutdown and transition 

activities (MMSC 1996). 

6.1.2 Purpose

This TBD represents a specific support mechanism for documentation of external dosimetry historical 

practices at the Pinellas Plant. It can be used to evaluate external dosimetry data for monitored 

workers and can serve as a supplement to individual monitoring data. For unmonitored workers, 

information in this document will provide for estimations of external doses. This document provides a 

site profile of Pinellas that contains technical basis information to be used to evaluate the total 

occupational external radiation dose for EEOICPA claimants.

6.1.3 Scope

This document provides supporting technical data to evaluate the total Pinellas occupational radiation 

dose that can reasonably be associated with the worker’s radiation exposure. This dose results from 

exposure to external radiation sources in Pinellas facilities that would be added to Plant 

occupationally required diagnostic X-ray examinations, and to onsite environmental releases, if 

applicable, to determine the total external dose. Also included are techniques to estimate the dose 

that might have occurred while an employee was not monitored or inadequately monitored; dose that 

might have been missed due to analytical detection limits; or dose for an employee whose monitoring 

records are incomplete or missing (i.e., missed dose). Over the years, new and more reliable 
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 10 of 59

scientific methods and protection measures have been deployed. The methods needed to account for 

these changes are also identified in this document.

This TBD describes the external dosimetry program at the Pinellas Plant. It discusses dose 

reconstruction, practices and policies at the Pinellas Plant, and dosimeter types and technologies for 

measuring dose from the different types of radiation present in the work environment. It also 

discusses the specific details of the evaluation of doses measured from exposure to electron (beta 

particles), photon (X-rays and gamma rays), and neutron radiation; sources of bias, workplace 

radiation field characteristics, responses of different dosimeters in the workplace radiation fields, and 

adjustments to the recorded dose measured by these dosimeters during specific years. 

6.1.4 Sources of Radiation

The manufacture of Pinellas Plant products required the use of radioisotopes. The predominant 

radionuclides used at the Plant included tritium, depleted uranium (DU), 85Kr, 238Pu, and 239Pu. In 

addition, the Plant used various analytical devices and calibration sources as part of the production 

process or for radiation detection device maintenance and calibration.

The predominant source of electron radiation at the Pinellas Plant was tritium, but this radionuclide 

emits only a low-energy beta particle with an average energy of 5.7 keV. Because electrons below 

15 keV do not have sufficient energy to penetrate the epidermal layer of the skin (NIOSH 2007), 

tritium is not considered to be an external radiation hazard. 

Starting in 1968, the Pinellas Plant stored tritium gas in beds of DU metal (MMSC 1994a, Phillips 

1975). DU metal in sealed stainless-steel containers stored tritium at normal pressure and 

temperature (GE 1979; Ward 1973; Weaver 1990). Heating can decompose the tritium bonds in 

uranium tritide to release ultrapure tritium while the remaining DU metal is ready for fast and 

reversible uptake of tritium at lower temperatures. The DU metal was fully contained in the tritium 

storage vessel. The storage vessels should not have been a source of external electron radiation 

exposure.

Krypton-85, a beta and gamma emitter, was used in two leak detection units as part of the Pinellas 

Quality Control Program. The leak detection units were housed in separate rooms and surrounded by 

ventilation shrouds. Each shroud was connected to ductwork that exhausted to the east main exhaust 

stack. Because it is a noble gas, 85Kr is a whole-body emitter with the possibility of a skin dose from 

electron radiation.

Small sealed sources of plutonium [consisting of 8.75 g and/or 10 g of source material, (GE 1982)]

produced at another site were shipped to the Pinellas Plant for use in the RTGs. The radionuclide 

composition of each sealed plutonium source was approximately 80% 238Pu, 16% 239Pu, 3% 240Pu, 

and 1% of other radionuclides by mass (GE 1982). The plutonium sources were enclosed in a triple 

metal encapsulation. These units were never opened. The capsules were inserted as heat sources 

for the RTGs (DOE 1983, pp. 2-11). The first recorded receipt of plutonium was on January 18, 1957, 

when the Plant received a 7-g (4.44 × 1012-Bq or 120-Ci) 238Pu source to calibrate Health Physics 

monitoring equipment. Seven plutonium heat sources arrived at Pinellas from Sandia National 

Laboratories on November 4, 1975 (GE undated a,b). About 50 RTG generators were produced 

every month (Burkhart 1987a, p. 4). The last onsite plutonium, with the exception of calibration check 

sources, was removed from the Plant in February 1991 (GE undated a,b).

Sources of ionizing radiation at the Pinellas Plant included other low-activity radioactive sources, such 

as those used to check or calibrate radiation detectors (i.e., calibration sources) and analytical 

devices employing radioactive byproduct material or X-rays produced by a radiation generating 

device. The calibration sources were maintained in the Health Physics Laboratory in Area 113 of 
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Building 100. While the exact inventory of radioactive sources varied over time, most were sealed 

microcurie check sources of radioactive isotopes such as 137Cs, 238Pu, 239Pu, 60Co, 14C, and 90Sr. 

These low-activity radioactive sources could have included alpha, electron, photon, and neutron 

emitters. For example, small quantities of 14C-labeled solvents were used in a laboratory testing 

operation. The largest 14C source was a 2.59 × 107

-Bq (700-μCi) source used for liquid scintillation 

counting calibration. Radiation doses from 14C resulted from internal deposition (DOE 1983). The 

analytical devices could have used radioactive byproduct material, such as 85Kr, 109Cd, or 241Am, that 

requires U.S. Nuclear Regulatory Commission certification. 

External sources of radiation at the Pinellas Plant included radiation generating devices. However, 

unlike radionuclide sources, radiation generating devices emit radiation only when they are turned on 

and activated. The most common type of radiation generating device at the Pinellas Plant was the 

primary product that it produced, which was the neutron generator. The neutron generators produced 

at the Plant were miniature accelerators that provided sources of fast neutrons for nuclear weapons. 

These miniature accelerators accelerated a beam of deuterium atoms into either a deuterium (D) or 

tritium (T) target (NCRP 1983). The resulting fast neutrons were produced by two different types of 

fusion reactions depending on the neutron generator design. Most neutron generators constructed at 

the Pinellas Plant produced 14-MeV neutrons via a T(D,n)4

He reaction (NCRP 1983; Weaver 1994a; 

Barclay 1995). However, a few neutron generators were constructed to produce 2.5-MeV neutrons 

via a D(D,n)3

He reaction (NCRP 1983; Weaver 1994a; Barclay 1995). These neutron generators also 

emitted X-rays from secondary reactions (NCRP 1983).

A full scale 200-keV Cockcroft-Walton ion accelerator was initially located in Room 161 of Building 

100 for about 2 years before being moved to Building 800 (Weaver 1994a; Barclay 1995). The 

accelerator was used to accelerate deuterium ions into various tritium targets to determine the 

neutron output of those targets (MMSC 1995a, p. 5-5). The accelerator was also used to irradiate 

battery compound samples (such as lithium and silicon) with high-flux neutrons (MMSC 1995b).

6.1.5 Occupational Dosimetry Program Overview

The Pinellas Plant started an external dosimetry program in 1957 to monitor individual employees 

working in neutron generator production areas. Some Pinellas records on facility monitoring, safety 

evaluations, investigations, etc., exist; however, most of these records concern operations after 1970. 

Records of radiation dose to individual workers from personnel dosimeters are generally available for 

1957 to 1994 for the workers’ time of employment. The dose measured by dosimeters was recorded 

at the time of measurement, reviewed by Pinellas health physicists, and routinely made available to 

workers. External Dose Reconstruction Implementation Guidelines (NIOSH 2007) indicates that 

these records represent the highest quality record for a retrospective dose assessment. The 

information in this section pertains to the analysis of the available records.

Table 6-1 lists the total number of Pinellas employees and the number of employees monitored for 

radiation exposure for years that this type of data was available. From 1960 to 1973, U.S. Atomic 

Energy Commission (AEC) annual exposure summary reports indicate that Pinellas had 27.5% of its 

labor force wearing dosimetry (377 of an average yearly labor force of 1,372). During the 1980s, 

while the data are not completely available, from 370 to approximately 400 of 1,650 to 1,975 workers 

(approximately 20%) were monitored for radiation dose. For the majority of Pinellas operations, 

external dosimetry was exchanged and analyzed monthly. Beginning in January 1990, external 

dosimetry was exchanged and analyzed quarterly (Burkhart 1988; Weaver 1990).
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Table 6-1. Personnel radiation dosimetry from AEC annual reports.

a

Year

Number of PAOAEC (or DOE) and 

GEND employees

Number of 

monitored 

employees

Number of employees 

in a given dose range

< 1 rem 1–2 rem

1960 1,304 225 225 0

1961 1,395 251 251 0

1962 1,370 254 254 0

1963 1,597 545 545 0

1964 1,408 347 347 0

1965 1,319 301 301 0

1966 1,445 325 325 0

1967 1,405 585 584 1

1968 1,397 281 281 0

1969 1,323 588 588 0

1970 1,311 442 441 1

1971 1,283 410 410 0

1972 1,402 346 346 0

1973 1,252 383 383 0

1974 NAb NA NA NA

1975 NA NA NA NA

1976 NA 317 317 0

1977 NA 300 300 0

1978 NA 297 297 0

1979 NA 334 334 0

1980 NA 376 376 0

1981 NA 389 389 0

1982 NA 408 408 0

a. Source: Data from Form AEC-190.

b. NA – not available.

No documentation was found to show that all employees were monitored at some time during Pinellas 

operations, and it is evident that not all groups of workers were required to wear dosimeters (ORAUT 

2005a). However, the Pinellas Plant was not required to monitor most employees because most of 

the work being performed did not expose personnel to external dose and because most personnel did 

not have routine access to radiation areas.

The external dosimetry assignment procedure has changed over the history of the Pinellas Plant. The 

earliest practices likely were based on regulatory requirements that required control of access, 

occupancy, and working conditions for radiation protection purposes (NBS 1958) and subsequently 

quarterly dose limits of 1.25 rem/quarter (AEC 1960). The AEC annual report information 

summarized in Table 6-1 illustrates that nearly all the personnel who were monitored for external dose 

were monitored when their annual exposures were less than 1.25 rem/yr. General operating 

procedures for the Pinellas Plant from as early as 1967 indicate that dosimeters were assigned to all 

personnel with a potential to receive a measurable external dose. This is confirmed by the annual 

whole-body doses that indicate that on average more than 79% of the monitored workers at the Plant 

received an annual whole-body dose of less than or equal to 0.02 rem for any given year (see Table 

B-1 in Attachment B). Given this information, any personnel with the potential to receive a 

measurable external dose at the Pinellas Plant were likely monitored for external dose. 

A GE general operating procedure titled “Film Badge Requirements,” dated May 10, 1967, states that 

it was management’s responsibility to “[a]scertain that film badges are worn by all personnel whose 

work assignments entail the potential for accumulating measurable radiation exposures; compliance 

with this procedure is a condition of employment in a Radiation Area” (GE 1957–1990). All versions 

of a GE general operating procedure titled “Assignment of Personnel to Work in Radioactive Material, 

Contamination, or Radiation Areas,” dated between January 17, 1968, and June 29, 1984, state that it 
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 13 of 59

was the Work Area Manager’s responsibility to “[e]nsure that all employees having work assignments 

in Radiation Areas are provided with film badges” (GE 1957–1990). In addition, a GE memorandum 

dated November 1, 1984, states that “[a]ll individuals with a potential of receiving measurable 

radiation dosage are included in our personnel monitoring program” (Greene 1984a). 

6.2 DOSE RECONSTRUCTION PARAMETERS

6.2.1 Historical Administrative Practices

Between 1957 and 1974, GEND-Health Physics conducted radiation dosimetry management and 

analysis through in-plant processing of X-ray and neutron-sensitive photographic films (Burkhart 

1987b). During 1965, GEND compared performance of GEND in-house processing results with 

commercial services from Landauer and Nuclear-Chicago for selected gamma and neutron doses, 

respectively, which showed essentially equivalent performance (Szedziewski 1965). In 1974, GEND 

contracted with R. S. Landauer Jr. & Co. (Landauer) to be the principal supplier and processor of the

dosimetry that GEND used. This arrangement continued until Pinellas Plant operations ended.

The available information indicates that, from the start of nuclear operations in 1957, the Pinellas 

Plant used film badge dosimetry available from or similar to the type sold by Nuclear-Chicago for the 

designated radiation control areas (GE 1957–1990). During the period that the Pinellas Plant read its 

own dosimeters, the available film work sheets indicate that only the open window (OW) film was read 

from the film dosimeters and the OW result was reported as photon dose (GE 1957–1990, pages 67 

and 99). This practice likely resulted in the photon doses reported for the period of 1957– June 1974 

being overestimates of the workers actual photon doses, because a significant portion of the Pinellas 

Plant’s photon doses were attributable to < 250 keV photons and because the OW portions of the 

early film dosimeters typically over-responded to the lower energy photons (ORAUT 2005c, 2006b). 

Because specific information on dosimeter design and dosimeter calibration could not be confirmed 

for the period of 1957– June 1974, no adjustments to the reported photon doses are recommended, 

to account for the over-response that the dosimeters likely had to the photon radiation at the Plant. 

The practice of only reading the OW film and reporting the result as photon dose would also explain 

why non-penetrating doses such as electron doses were not reported during this era. In addition, any 

potential electron doses received by the workers would have been accounted for and reported as 

photon doses by this practice. 

An unsigned note in a collection of highly varied examples of dosimetry records, which include a 

number of records from the GEXM site, indicates that personnel monitoring for neutrons might not 

have began at the Pinellas Plant until 1960, and neutron doses for the period from 1957 through May 

1960 might have been estimated from area monitors (GE 1957–1990). Even if the note is accurate 

about neutron monitoring at the Plant, it should not have an adverse effect on the neutron dose 

calculations for that period. Because the same whole-body dosimeters were used for personnel and 

area monitoring, the neutron dose calculations would be the same regardless of whether the worker’s

neutron doses were based on a personnel dosimeter or a dosimeter used to monitor an area.

Beginning in July 1974, the Plant began using dosimetry provided exclusively by Landauer. The 

original Landauer dosimetry was based on film badge technology (Ward 1974). Landauer processed 

the dosimetry and provided exposure reports to GEND for review after badge processing. Figure 6-1 

is a replication of a portion of a Landauer dosimetry report (Landauer 1986-1988). Information in the 

Landauer reports included personnel data (identification number, name, and social security number), 

dosimeter type, deep and shallow exposure for the reporting period (i.e., monthly), and cumulative 

totals for deep and shallow exposures for the calendar quarter, year to date, and permanent exposure 

(lifetime exposure at Pinellas). The exposure information was entered in each person’s exposure 

history by hand or into a Pinellas-based computer system and GE’s Corporate Health and Safety 

Record (HSR) system (Richards 1986). 
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 14 of 59

Starting in mid-1978, the Plant began using the Landauer polycarbonate plastic dosimeter for 14-MeV 

neutrons, and continued to use photographic film processing for 2.5-MeV neutrons, X-ray, beta, and 

gamma exposures, and Landauer thermoluminescent dosimeter (TLD) rings for hand monitoring 

(Burkhart 1987b). 

From October 1979 through September 1987, the Plant used dosimeters from Mound Laboratory for 

evaluating exposures to X-rays and 2-MeV average neutrons from the handling of sealed 238PuO2

sources during the production of RTG units in Building 400, and continued to use the Landauer 

dosimetry discussed above for exposures from all other radiation sources, including 14-MeV neutrons. 

Problems related to the DOE Laboratory Accreditation Program (DOELAP) testing with the Mound 

dosimetry and the equivalent performance of Landauer neutron dosimetry led GEND to discontinue 

use of the Mound dosimetry in October 1987 (Burkhart 1987b). Beginning in 1990, earlier dosimetry 

technology was replaced with Landauer TLD dosimetry that was used until the end of nuclear 

development and testing operations in 1994. Table 6-2 summarizes the significant historical events 

for the dosimetry program at the Pinellas Plant.

Table 6-2. Pinellas Plant historical dosimetry events.

Date Event Reference

1957 In-plant processing of gamma and neutron film dosimeters initiated. Burkhart 1987a

April 1957 New employee orientation in radiation safety offered. Burkhart 1990

October 1957 Measured neutron dose rates at all test positions. Burkhart 1990

November 1957 Measured neutron output of neutron generators, and was 

determined to be equivalent to 10 mrem/pulse at 1 in.

Burkhart 1990

December 1957 Sandia National Laboratories (SNL) asked to provide film badges.

However, there is no indication that SNL ever provided any film 

badges for the Pinellas Plant.

Burkhart 1990

January 1960 Full-time Health Physics representative assigned to Area 108. Burkhart 1990

November 1963 Began use of wrist badges in place of ring badges for limited 

number of employees.

Burkhart 1990

February 1965 GEXM memorandum comparing performance of two types of 

neutron badges and two types of gamma badges.a Szedziewski 1965

Late 1969 Film badge fading study. Author unknown

1969

January 1973 Memorandum on dose rate for stress test facility. Holliday 1973

July 1974 Began using Landauer as source of film badges. Ward 1974

April 1978 Memorandum on personnel neutron dosimetry recommending use 

of new Landauer neutron badge using polycarbonate plastic.

Holliday 1978

October 1979 Began using Mound neutron dosimeters. Burkhart 1987b

1986 Landauer switched from reporting photon doses as exposures to 

reporting them as deep doses.

Yoder 2005

October 1986 Memorandum on estimated doses to GEND personnel handling 

unmarked neutron generator units.

Burkhart 1986

October 1987 End use of Mound neutron dosimeters. Burkhart 1987b

October 1988 Memorandum on radiation dose rates from RTG heat sources. Weaver 1988

April 1990 Changed from Landauer film to thermoluminescent dosimetry. Hall 1989

1971–1993 Various determinations for doses from testing of sealed neutron 

generators.

GE 1971-1996

September 1994 Nuclear product development and testing end. Other radiationrelated work did continue. 

a. The Pinellas Plant was the progeny of General Electric’s X-ray Division in Milwaukee (GEXM). 

Because the radiological control programs for these two sites were closely linked and because the 

available information indicates that the Pinellas Plant managed the radiological programs at GEXM 

after its startup, the information in this evaluation is considered to be applicable to the Pinellas 

Plant dosimeters.
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 15 of 59

Attachment A contains examples of handwritten and computer records for 1960 to 1988. The forms 

used in 1960 were also used before 1960 and the forms used in 1988 were used until Pinellas ceased 

operations.

A Pinellas Plant health physicist reviewed the reports and evaluated and resolved unusual or 

inconsistent results. The health physicist could modify the reports, documenting all investigations and 

reasons for such modifications. These reports were placed in the worker’s dosimetry file. The health 

physicist checked the printed version of the Landauer reports against the electronic version. Until 

1990, workers who reported lost or stolen badges were assigned an exposure that was an average 

from their previous dose histories (GE 1990a).

If Landauer found that a badge exceeded 400 mrem whole body, 800 mrem skin, or 6,000 mrem 

extremity, it was required to call the responsible Pinellas health physicist (GE 1990a). Analysis of the 

available claimant records found no documentation that this reporting requirement was ever exceeded 

during the time Landauer provided dosimetry services to the Plant (1974 to 1994).
Document No. ORAUT-TKBS-002

9

6 Revision No. 01 Effective Date: 04/28/2011 Page 16 of 59

Figure 6-1. Regenerated example of a Landauer dosimetry report from the 1970s.

GENERAL ELECTRIC COMPANY ACCOUNT NO. 70463

NEUTRON DEVICES DEPT

ENVIRONMENTAL HLTH

7887 BRYAN DAIRY RD

LARGO FL 33543

Participant 

ID 

number Name

Social 

security 

number

Note 

(see 

reverse 

side)

Dosimeter 

type Use

Radiation 

quality

Exposure to 

badge (millirems) 

for period(s) 

indicated below

Cumulative totals 

(millirems)

Adjustments Sex

Birth date

Number 

badge reports

Inception 

date

Calendar 

quarter

Year to 

date Permanent

Deep Shallow Deep Shallow Deep Shallow Deep Shallow MO DA YR To date Qtr MO YR

[PIR] AREA MONITOR 183 E 1 M M M M M M M M 11 1 11

[PIR] [PIR] [PIR] U 3 M M M M 24 2 11 84

[PIR] CONTROL G 1 M M M M M M 260 260 295 5 6 74

[PIR] TLD CONTROL M M M M 101 3 2 79

[PIR] LOAN U 3 M M M 220 96 3 2 79

[PIR] [PIR] [PIR] G 1 M M M M M M M M 7 5 9 86

[PIR] [PIR] [PIR] G 1 M M M M M M 100 100 M 256 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M M M 250 250 M 256 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M M M 110 110 F 256 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M M M 10 10 23 3 11 84

[PIR] [PIR] [PIR] G 5 M M M M 23 3 2 85

[PIR] [PIR] [PIR] G 1 M M M M M M 240 240 M 256 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M 30 30 250 250 M 255 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M 40 40 90 90 M 256 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M M M 90 90 M 256 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M M M 200 100 M 256 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M M M 170 170 M 257 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M M M 180 180 M 258 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M 40 40 140 140 M 254 5 6 74

PIR = privacy information removed.
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 17 of 59

6.2.1.1 Performance Testing

Neutron Studies

Eastman Kodak nuclear emulsion Type A (NTA) film was used for neutron measurements and should 

be similar to the Landauer NTA dosimetry used at Pinellas after 1974 for the Type J dosimeter 

(Author unknown 1969; Gordon 2004). As stated in the Hanford Site, Idaho National Laboratory, and 

Nevada Test Site Occupational External Dose TBDs (ORAUT 2010a, 2007, and 2010b, respectively), 

NTA was basically the only common dosimeter method available to measure neutron dose in AEC 

facilities prior to the use of thermoluminescent dosimeters (TLD). The neutron spectra at Pinellas 

were known to be dominated by higher energy 14-MeV deuterium-tritium and 2.5 MeV deuteriumdeuterium fusion neutrons due to the unique designs of the neutron generators. These higher-energy 

neutron fields tend to minimize errors from fading or energy response reported at the first AEC 

Neutron Dosimetry Workshop in 1969 which indicated that Savannah River Site calibration laboratory 

dose measurements made with NTA film were about one-half to one-fourth of those measured with 

other methods, including the neutron TLD (Vallario et al. 1969). 

Typically, the dominant neutron exposure from nuclear weapon components at Pinellas was readily 

and reliably measured with NTA film dosimeters. For neutrons from RTG production, Pinellas initially 

continued with NTA film but changed to polycarbonate dosimeters beginning in June 1978. From 

October 1979 through September 1987, TLDs from Mound Laboratory were used for measuring 

neutron exposures from RTG production. From September 1987 through the end of RTG production, 

Mound Laboratory dosimeters were out phased with a Landauer Type R dosimeter arrangement 

(Burkhart 1987a).

Specific Pinellas Plant Studies

The NTA film badge neutron dosimeters used from 1957 through 1977 underwent track fading during 

use, causing a loss of information (Holliday 1978). At least for a period, the neutron dosimeters were 

made of Kodak Type A film (Author unknown1969). GEND established a factor of 3 to correct for 

track fading beginning in January 1970, based on a track fading study (Author unknown1969). The 

maximum errors associated with the use of this factor occurred when a worker received a majority of 

the neutron exposure at the beginning or the end of a dosimeter monitoring period. The assigned 

dose to a worker receiving a total exposure on the first day of a dosimeter period would be 20% of the 

true dose per week; while the dose assigned to a worker receiving a total exposure on the last day of 

the monitoring period would be 40% of the true dose for a monthly wear period (Author 

unknown1969). The Landauer polycarbonate badge that replaced these badges in May 1978 did not 

undergo track fading, so no correction factor was applied after May 1978 (Holliday 1978). The 

Landauer Type I dosimeter, which was used in the Type R dosimeter package, had a limited 

sensitivity to neutrons from the sealed PuO2 sources in the RTG units; this was calculated to be 67% 

of the true dose by GEND health physics personnel. 

In 1965, a dosimeter comparison study was performed. Part of the study compared the neutron 

dosimeters being processed at the Pinellas Plant to dosimeters obtained from and processed by 

Landauer. The study concluded that the neutron dosimeters processed by the Pinellas Plant were 

preferred over the Landauer dosimeters for neutron generator work (Szedziewski 1965).

6.2.1.2 Reports

An As-Low-As-Reasonably-Achievable (ALARA) Program report discussing occupational exposures 

was submitted to the manager of the Environmental Health and Safety Program. Attachment A 

contains examples of the reports kept by the Environmental Health and Safety Program and its 

radiation protection program from 1960 to 1988 for one claimant.
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 18 of 59

6.2.2 Dosimetry Technology

For the period of 1957–June 1974, no information regarding the specific design parameters for the 

film dosimeters used at the Pinellas Plant (e.g. film type, filters used, dosimeter holder design, etc...) 

could be found. There is limited documentation that indicates there might have been an early 

relationship with Nuclear-Chicago (GE 1957–1990), but that relationship might have only been with 

the Pinellas Plant’s parent site (i.e., GE’s X-ray Division in Milwaukee [GEXM]). Nothing has been 

found to confirm the use of the Nuclear-Chicago film dosimeters or dosimeter holders at the Pinellas 

Plant. For the period after June 1974, the dosimetry type and sources are well documented in the 

Landauer exposure reports. Table 6-3 summarizes the monitoring technique and describes 

suspected and known Plant dosimeters. Tables 6-10 and 6-11 in Section 6.4 summarize the limits of 

detection (LODs) of these dosimeters along with the maximum potential missed doses (NIOSH 2007).

Table 6-3. Dosimetry used at Pinellas Plant for external whole-body and extremity exposuresa

.

Period Dosimeter type Dosimeter description

Whole body – beta-gamma dosimeters

1957–Jun 1974 Photographic film

(beta-gamma) 

Nuclear-Chicago or similar film dosimeters processed in-house 

(Burkhart 1987b). Nuclear-Chicago film dosimeters for this era 

contained a single film packet. Three filters consisting of cadmium, 

aluminum, and lead were incorporated into fronts and backs of the 

film dosimeters for energy dependence (ORAU 2003). However, it 

could not be confirmed that these were the dosimeters used at the 

plant.

Jul 1974–Mar 

1990

Landauer Type Gb,c

(beta-gamma)

The Type G dosimeter was a film emulsion package placed in 

standard Gardray holder/badge for monitoring beta, X-ray, and 

gamma exposures. It was insensitive to neutron radiation. 

Required in areas where Kr-85 was used. Required for radiationgenerating equipment and accelerator operators.

Gamma and X-ray: 30 keV to 20 MeV; beta: over 1.5 MeV.

Apr 1990–1997 Landauer Type Z/Fc

(beta-gamma)

The Type Z dosimeter was comprised of 3 TLD-700 chips for 

monitoring beta, X-ray and gamma exposures. It was insensitive to 

neutron radiation. After 1994, the Type Z dosimeter was renamed 

the Type F dosimeter.

Whole body – neutron and beta-gamma-neutron dosimeters

1957– Jun 1974 NTA film

(neutron only)

Nuclear-Chicago or similar film dosimeters that were equipped with 

Eastman/Kodak Nuclear Track Type A (NTA) film emulsions and 

processed in-house (Burkhart 1987b). Fast neutrons undergoing 

elastic collision with content of emulsion or cellulose acetate base 

material produced recoil protons, which were recorded as 

photographic tracks in the NTA film emulsion. Track density was a 

linear function of dose. Developed images exhibited tracks caused 

by neutrons, which could be viewed using appropriate imaging 

method (i.e., oil immersion) and 1000X-power microscope or 

projection capability. 

Jul 1974–May 

1978

Landauer Type Pb,c

(beta-gammaneutron)

The Landauer Type P dosimeter appears to have been used in all 

areas with a potential for neutron exposures (Landauer 1974–

1980). The Type P dosimeter was a combination beta-gamma and 

fast neutron dosimeter. The fast neutron dosimeter was an NTA 

film dosimeter and the beta-gamma dosimeter was likely a 

Landauer Type G film dosimeter. 
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 19 of 59

Period Dosimeter type Dosimeter description

Jun 1978–1997 Landauer Type Ec

(neutron only)

The Landauer Type E neutron dosimeter was used in all areas at 

the Pinellas Plant until October 1979 when the site started to use 

the Mound dosimeters in areas with RTG operations. After October 

1979, the Type E dosimeter was used only in the neutron generator 

areas at the Plant. 

The Type E dosimeter is a polycarbonate (Lexan®) neutron recoil 

track registration device used to monitor fast neutron interactions. 

Neutrak 144 has a dosimeter element for response to fast neutrons. 

Neutrak E1 has a polyethylene radiator over CR-39 chip that would 

monitor for fast neutrons; only Lexan® responded to neutrons by 

recording ionization damage caused by neutrons interacting with 

carbon and oxygen atoms, which leaves a track. It had uniform 

energy response from 3 to over 14 MeV with threshold of about 

1 MeV (Weaver 1987). 

The Type E dosimeter was combined with the Type G dosimeter 

(and later with the Type Z/F dosimeter) (Weaver 1987, 1996). 

Workers were required to wear an E/G dosimeter combination when 

working around neutron generators. Accelerator operators were 

also required to wear an E/G dosimeter combination. An E/G 

dosimeter combination or a G dosimeter was required when 

working with calibration sources (Weaver 1987). 

Oct 1979–Sep 

1987d

Mound TLD

(gamma-neutron)

The Mound TLD was used in areas with RTG operations (i.e., 

where PuO2 was handled). It was used to measure exposures to 

the X-rays and 2-MeV average neutrons from the handling of the 

sealed PuO2 sources during the production of RTG units in Building 

400. This dosimeter utilized a Harshaw 8810 TLD package that 

included an albedo neutron monitoring capability. The Harshaw 

8810 TLD package utilized a combination of TLD-600 and TLD-700 

dosimeter chips, which were encased in a plastic holder made of 

Cycolac®.

Sep 1987d–Feb 

1991

Landauer Type R/Ic

(beta-gammaneutron)

The Landauer Type R dosimeter arrangement was used in areas 

with RTG operations (i.e., where PuO2 was handled), and replaced 

the Mound TLD. The Type R dosimeter package consisted of a 

Type G dosimeter for measuring beta, X-ray, and gamma radiation 

and a Type I dosimeter for measuring neutron radiation. The Type I 

dosimeter combined a TLD albedo neutron monitor with a track 

recoil device (CR-39 [allyl diglycol carbonate]) that responds to 

proton recoil events. Neutron energy range was approximately 

1 x 10-6 to 10 MeV. Albedo response to thermal neutron radiation 

was subtracted to yield fast neutron dose. The “Neutrak ER” has an 

albedo element with above-described elements. Qualitative 

relationship was derived to determine ratios of neutrons of various 

energies. When first used at the Pinellas Plant, the Type I 

dosimeters did not meet all DOELAP requirements during 

performance testing (Weaver 1991). 

After the Type G dosimeter was replaced by the Type Z dosimeter, 

the combined dosimeter unit was known as the Type I dosimeter. 

After production of RTGs was halted in October 1991, the Type R/I 

dosimeter was used only as an area monitor for the americiumberyllium (AmBe) source (Weaver 1991).
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 20 of 59

Period Dosimeter type Dosimeter description

Wrist – beta-gamma dosimeters

Jul 1974–1990 Landauer Type Gc,e

(beta-gamma)

The Type G wrist dosimeter was a film emulsion package placed in 

a standard Gardray holder/badge for monitoring beta, X-ray, and 

gamma exposures. It was insensitive to neutron radiation. The 

Type G wrist dosimeter was used to monitor extremity doses in 

areas with RTG operations (i.e., where PuO2 was handled) (Weaver 

1987); extremity dosimetry was assigned in other areas as needed 

(Weaver 1996).

1991–1997 Landauer Type Kc

(beta-gamma)

The Landauer Type K wrist dosimeters assigned during this period 

utilized three TLD-100 chips.

Finger ring – beta-gamma dosimeters

1957–1974 Photographic film

(beta-gamma) 

Nuclear-Chicago or similar film ring dosimeters processed in-house 

(Burkhart 1987b). Nuclear-Chicago film ring dosimeters contained 

a single film packet. Three filters consisting of cadmium, aluminum, 

and lead were incorporated into fronts and backs of the film 

dosimeters for energy dependence (ORAU 2003).

1975–1983 (unknown) This analysis found no information regarding the type of ring 

dosimeters used during this period. If ring dosimeters were used 

during this period, they might have included the in-house film 

dosimeters or Landauer Type L ring dosimeters. Because the type 

of ring dosimeter that might have been used is uncertain, extremity 

doses based on ring dosimeters from this period will need to be 

assessed on a case-by-case basis.

About 1983f–1997 Landauer Type Uc

(beta-gamma)

The Landauer Type U ring dosimeters assigned during this period 

utilized one LiF TLD chip for monitoring beta (>1.5 MeV), X-ray, and 

gamma exposures. It was insensitive to neutron radiation. The 

Type U ring dosimeter was used to monitor extremity doses in 

areas with RTG operations (i.e., where PuO2 was handled) (Weaver 

1987); extremity dosimetry was assigned in other areas as needed 

(Weaver 1996).

a. Sources: Battelle 1956–2002; Burkhart 1987b; GE 1986–1988, 1990a,b; Greene 1985a; Hall 1989; Holliday 1978; 

Landauer 1974–1980, 2004; Ingle 1991; Weaver 1987, 1991, 1995, 1996; Passmore 2004; Gordon 2004.

b. A June 1974 memorandum indicates that the Pinellas Plant ordered Landauer Type J (beta-gamma) and Type K (betagamma and neutron) dosimeters (Burkhart 1987b); however, the actual dosimetry records indicate that Landauer Type 

G (beta-gamma) and Type P (beta-gamma and neutron) dosimeters were actually received and used (Landauer 1974–

1980). 

c. The use of the Landauer dosimeters is indicated by exposure type codes in the earlier years and use codes in the later 

years. The following are the dosimeter exposure type or use codes: 1 - whole body, 2 – skin/lens of eye, 3 - right finger, 

4 – left finger, 5 – right wrist, 6 – left wrist, 7 – other extremity, 8 – other whole body, 9 - monitor.

d. Dual Mound and Landauer Type R neutron dosimeters were used to monitor RTG workers during the month of 

September 1987 (Burkhart 1987b).

e. A June 1974 memorandum indicates that the Pinellas Plant ordered Landauer Type M (beta-gamma) wrist dosimeters 

(Burkhart 1987b); however, the actual dosimetry records indicate that Landauer Type G (beta-gamma) wrist dosimeters 

were actually received and used (Landauer 1974–1980).

f. This analysis found no documentation that shows the start of use for the Landauer Type U finger ring dosimeters.

6.2.3 Dosimetry Calibration Practices

Pinellas conducted dosimetry calibration as part of its external dosimetry audit program. This type of 

performance testing occurred every 6 months using radiation sources with known strengths. Tests 

used the DOE Standard for the Performance Testing of Personnel Dosimetry Systems (DOE 1986) 

and American National Standards Institute Personal Dosimetry Performance – Criteria for 

Performance (ANSI 1993). Each test used approximately six to nine badges.

The photon calibration of the Landauer Type G dosimeters was performed by exposing the 

dosimeters to the Shepherd model 81-12 137Cs beam irradiator in Building 800 (a gamma check for 

662-KeV photons). For beta calibration, the dosimeters were exposed by placing them on a bare 
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 21 of 59

uranium slab for exposure to the resulting radiation. A covering with a known density thickness was 

placed on the dosimeters to keep them free from uranium contamination (GE 1990a).

Calibration of the Landauer Type E polycarbonate dosimeters was performed by exposing them to a 

D-T fast neutron source with known source strength. The badges were placed on a Lexan® “jig” and 

set at a known distance from the source of neutrons (GE 1990a).

The Landauer Type R dosimeter, which consisted of a beta-gamma film dosimeter, an albedo neutron 

monitoring TLD, and a CR-39 neutron dosimeter, was placed on a water phantom and exposed to a 

Shepherd Model 149 calibrator equipped with an 241Am-Be source. The phantom was level with (and 

at known distances from) the source on a moveable metal rack about 4 ft above the floor to minimize 

scattering effects (GE 1990a).

However, even though there is Pinellas Plant documentation showing such calibration studies 

occurred, the results of the studies and subsequent use in the radiation dosimetry program are not 

available.

Beginning in 1974, Landauer supplied all dosimetry badges and performed the necessary calibrations. 

Landauer used control film. The personnel monitoring reporting was normally in net exposure; the 

control film reading was deducted from the personnel film reading. If the control film appeared to 

have been exposed differently than the personnel packets, the densities on the personnel film were 

normalized to Landauer controls only and a nonminimal control reading was reported. A control 

packet reading was provided in arbitrary units, not necessarily in millirem. Minimal beta or soft X-ray 

skin dose readings were not reported until after a positive skin dose exposure was recorded. Ring 

badges were calibrated only for high-energy gamma (probably > 0.662 MeV) and high-energy beta 

(1.5 MeV) unless special arrangements were made with the Plant (Gordon 2004).

Further details of the dosimetry are listed in Tables 6-9 and 6-10 in Section 6.4. This includes the 

dosimeter types along with some dosimeter configuration and energy response characteristics. 

Table 6-9 lists the associated LODs and the maximum potential missed photon and electron doses. 

Table 6-10 lists the associated LODs and the maximum potential missed neutron doses. Even though 

the dosimeter designations or types changed, the LODs did not vary much over the operational 

history of the Pinellas Plant.

6.2.4 Workplace Radiation Fields

Potential sources for workplace radiation fields at Pinellas can be placed in two categories, 

radionuclides and machine-generated X-rays, neutrons, and electrons. The only open-area radiation 

fields to be routinely encountered by workers would be from the testing of neutron generators, the use 

of machine-generated X-rays, and areas where work was performed with the RTG heat sources. 

Radionuclides were used directly in the manufacturing processes, manufacturing support, and various 

calibration sources. Table 6-4 lists radionuclides that would contribute to workplace radiation fields. 

The radionuclides applied in the manufacturing processes at the Pinellas Plant were kept in 

containers, in sealed sources, or in the process piping. Because most of the primary radiation from 

these sources (alphas and betas) would not penetrate the containers, sealed source encapsulation, or 

the process piping, and the workers would not be in close contact with the sources, the expected 

radiation fields would be very small. In the cases of tritium and 85Kr, which were gases in the Pinellas 

processes, vent hoods or direct connections to ventilation exhaust systems quickly removed the 

gases from worker spaces. The largest calibration source was a 120-Ci 137Cs source used in a Model 

81-12 Beam Calibrator manufactured by J. L. Shepard and Associates (GE 1977). Three neutron 

sources were used at various times, two of which contained 239Pu (1.7 Ci and 0.43 Ci) and an Am-Be 

source with 10 Ci of 241Am. All three were removed from the Pinellas site by the end of 1991 (1979, 
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 22 of 59

1990, and late 1991, respectively). All other calibration sources were in the millicurie and microcurie 

ranges and were sealed, and would not contribute to occupational doses (other than to workers who 

used the millicurie sources) (GE undated b, pp. 6-10).

There were also machine-generated radiation fields at the Pinellas Plant. X-rays and neutrons were 

generated from the neutron generators and X-rays were generated from other equipment listed in 

Table 2-1 of the Technical Basis Document for the Pinellas Plant – Site Description (ORAUT 2011b). 

For neutron generator tests, there was no shielding around the test area with the exception of the 

room walls. X-ray-generating equipment was designed to be in a shielded enclosure that, by 

certification, was below 0.5 mR/hr at 5 cm from the surface of the device. Most devices would have a 

lower radiation field. Shielding around the test area was added later at an unspecified time. There 

was also X-ray production from the tritium tube testing. This is implied by a description of exposures 

in a GE memo (Greene 1984a) and in a memo from Sandia National Laboratories describing X-ray 

generation from neutron tube testing (Brainard 1991). By calculation, it is assumed that X-ray dose 

was linear to the beam current of electrons and ions. The dose equivalent of X-rays produced was 

close to that of the neutron dose equivalent without any shielding of the tube present (Brainard 1991).

To support the production of RTGs, Pinellas health physicists measured RTG radiation fields to yield 

3 mrem/hr gamma and 10 mrem/hr neutron at 10 cm (approximately 4 in.) from the surface of the 

device and measurements during a December 17, 1987, hazard review yielded 20 mrem/hr gamma 

and 60 mrem/hr neutron at contact. Both measurements were most likely from one RTG unit (GE 

1987). 

Another indicator of the expected radiation fields experienced by Pinellas workers comes from a 

review of the dosimetry records for the first 285 Pinellas Plant claims in the NIOSH Office of 

Compensation Analysis and Support Claims Tracking System (NOCTS) database (i.e., up through 

Claim No. [redacted]). Of the 285 claims, 60 workers had whole-body doses greater than 0.001 rem 

from all forms of radiation, with a cumulative dose of 16.823 person-rem. Of this cumulative dose, 

7.152 person-rem (42.5%) was from external neutron exposures, 6.587 person-rem (39.2%) was from 

external beta/gamma exposures, and 3.085 person-rem (18.3%) was from internal tritium exposures. 

Of these 60 workers, 32 have lifetime whole-body doses greater than 0.100 rem, and 9 have Pinellas 

lifetime doses from all sources greater than 0.500 rem. Most of these doses were accumulated in a 

sporadic manner by receiving a dose over a fraction of a year and/or not having any recordable doses 

for several months or for years in some cases. Because a limited number of monitored employees 

received extremity dosimetry, only 5 of the 60 workers with external doses have recorded lifetime 

extremity doses. The extremity doses for those individuals were 0.020, 0.028, 0.056, 0.483, and 

0.730 rem. The exposure histories in these 285 claims are supported by the annual radiation 

exposure summary reports compiled by the AEC from 1960 to 1973, which indicates that there were 

only two instances when a Pinellas Plant worker received more than 1.0 rem in a given year.
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 23 of 59

Table 6-4. Radionuclides contributing to workplace radiation fields.

Nuclide Source Half-life

Energies and abundances of major radiationsa

Alpha (MeV)

Average

beta (keV)

Photon 

(keV)b Neutron (MeV)

Kr-85 Leak check 

source

10.72 yr --- 251.4 

(99.6%)

514 (0.43%) ---

Cs-137 Calibration 

source

30.17 yr --- 156.8 

(94.6%)

415.2 

(5.4%)

661.6 

(89.98% from 

decay of 

Ba-137m)

---

Am-241 Part of Am-Be 

sealed neutron 

source

432.2 yr 5.39 (1.4%)

5.44 (12.8%)

5.49 (85.2%)

--- 26.3 (2.4%)

59.5 (35.9%)

Thermal neutrons by a 

9

Be(α,n) reaction

DU metal powder for H-3 storage

U-238 99.75% weight 4.47E9 yr Not listed. --- --- By weak spontaneous fission 

U-235 0.25% weight 7.04E8 yr Not listed. --- 109.1 (1.5%)

143.7 (10.5%)

163.4 (4.7%)

183.7 (54%)

202.1 (1.0%)

205.3 (4.7%)

By weak spontaneous fission

U-234 0.0005% 

weight

2.45E5 yr Not listed. --- --- By weak spontaneous fission 

Radionuclide composition of sealed PuO2 sources used for RTG unitsc

Pu-238 80.2% weight 87.75 yr 5.46 (28.3%)

5.50 (71.6%)

--- --- By weak spontaneous fission 

and 18O(α,n) reaction 

Pu-239 15.9% weight 24,131 yr 5.10 (11.5%)

5.14 (15.1%)

5.16 (73.3%)

--- --- By weak spontaneous fission 

and 18O(α,n) reaction 

Pu-240 3.022% weight 6,569 yr 5.12 (26.4%)

5.17 (73.5%)

--- --- By weak spontaneous fission 

and 18O(α,n) reaction 

Pu-241 0.643% weight 14.4 yr --- 5.2 (100%) --- ---

Pu-242 0.132% weight 3.76E5 yr 4.87 (22.4%)

4.90 (78.0%)

--- --- By spontaneous fission and 

18O(α,n) reaction

U-234 0.140% weight 2.45E5 yr 4.72 (27.4%)

4.78 (72.4%)

--- --- By weak spontaneous fission 

and 18O(α,n) reaction

Np-237 0.130% weight 2.14E6 yr 4.64 (6.2%)

4.66 (3.3%)

4.71 (1.0%)

4.77 (33.0%)

4.79 (47.0%)

4.80 (1.6%)

4.81 (2.5%)

4.87 (2.6%)

--- --- By weak spontaneous fission 

and 18O(α,n) reaction

Am-241 0.0033% 

weight

432.2 yr 5.39 (1.4%)

5.44 (12.8%)

5.49 (85.2%)

--- 26.3 (2.4%)

59.5 (35.9%)

By weak spontaneous fission 

and 18O(α,n) reaction 

a. Energy information from Kocher (1981); with the exception of Kr-85, radiation emissions with abundances less than 

1.0% were not listed in this table; the alpha particles are not an external dose hazard and are only being listed in this 

table when they are a significant source of neutrons via an alpha-neutron reaction (α,n).

b. The characteristic K and L shell X-rays have not been included in this table because the encapsulation materials for the 

affected sources would have attenuated their emissions to an insignificant level.

c. Radionuclides that are present in quantities less than 0.001% by weight are not listed; the encapsulation material for 

these sources would have attenuated the photon emissions from the ≤ 60 keV photons to less than 1/20,000th of their 

original intensity; also, the photons emitted from the spontaneous fission reactions, alpha-neutron reactions, and decay 

of fission and activation products are not listed in this table.
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 24 of 59

6.2.4.1 Electron and Photon Exposures

The NIOSH Interactive RadioEpidemiological Program (IREP; NIOSH 2007) discusses three photon 

energy bands – below 30 keV, 30 to 250 keV, and above 250 keV. The principal source of photons to 

the greater number of Pinellas workers would have been the various X-ray generating devices due to 

the limited access by a small number of workers to direct contact with radionuclides that emit photons 

(

85Kr in the two leak detection systems, the one 137Cs sealed source in a shielded calibration machine, 

DU in the tritium storage beds, and 238PuO2 RTG heat sources). 

The X-ray machines would generate photon energies below 250 keV. Because the machines were 

shielded, those used probably have a range of energy settings; and because measurements of the 

photon energies on the outside of the shielding are not available, this analysis grouped this source of 

photons in the 30- to 250-keV energy interval, which should be favorable to claimants. Table 6-5 lists 

electron and photon energies and percentages for processes at the Pinellas Plant.

Table 6-5. Electron and photon radiation energies and percentages for Pinellas Plant processes.

Facility or location Process type Radiation type

IREP energy 

group Percentage

Buildings 100, 200, 

300, 400, Medical

Neutron generator production, 

quality assurance, and RTG 

production

Photon 30–250 keV 100

Building 100, Area 

109

Product analysis (Radiflo® and 

X-ray)

Electron

Photon

> 15 keV

30–250 keV

> 250 keV

100

86

14

Building 800 Calibration and accelerator Photon 30–250 keV 100

External electron exposures due to beta particles were unlikely. The predominant source of electron 

radiation at the Pinellas Plant was tritium, but this radionuclide emits a low-energy beta particle that 

averages only 5.7 keV. Because electrons below 15 keV do not have sufficient energy to penetrate 

the epidermal layer of the skin (NIOSH 2007), tritium is not considered an external radiation hazard. 

The only sources of electron radiation with sufficient energy to penetrate the skin are 85Kr and 137Cs. 

(Note: There was the use of DU beds for tritium processing, but the DU was in small metal cylinders 

that would have attenuated the beta component from the worker.) Several incidents occurred in 

which workers were exposed to 85Kr, so potential exposures to electron radiation should be 

considered in the area where the two 85Kr leak detection systems (Radiflo® and TracerFlo systems)

were located. The 137Cs was in a sealed source in a shielded cabinet in the concrete Building 800, so 

the probability of a worker receiving electron radiation exposures from 137Cs was remote. 

Electron radiation exposures were possible for X-ray diffraction and electron beam devices if 

containment of the beams was compromised. It is more likely that exposures from these devices

would have been from X-rays or bremsstrahlung production and not from any free electron beam. 

The exposures, if diffuse, would have been monitored by film badge or TLD. Most unusual 

occurrences were documented and any likely penetrating or nonpenetrating exposures would have 

been addressed in personnel medical records. 

6.2.4.2 Neutron Exposures

There were two distinct neutron sources at the Pinellas Plant – 2.5- and 14-MeV neutrons from the 

testing of neutron generators and 2-MeV average neutrons from the sealed plutonium sources in the 

RTGs. The locations of the operations producing neutrons include a small number of areas. 

However, personnel wore neutron dosimetry in other locations; Table 6-6 lists the locations where 

workers wore neutron dosimetry and the expected neutron source energies.
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 25 of 59

Table 6-6. Expected neutron radiation energies for Pinellas Plant 

processes.

Facility or 

location Process type

Neutron energy

(MeV)

107 Tube assembly 2-14 

128 Tube test 2-14 

131 Final test 2-14 

132M Fan room 2-14 

182 Tube assembly 2-14 

183 General development 2-14 

184 Tube testing 2-14 

191 CPE hood room 2-14 

194 Engineering environmental testing 2-14 

Radioanalytical Laboratory 2-14 

Building 200 Product QA testing 2-14 

Building 300 2-14 

Building 400 RTG assembly and testing 0-12 

Building 800 Calibration and accelerator 2-14 

The RTG PuO2 heat source spectrum is illustrated in Figure 6-2 (Greene 1984b). The spectrum was 

probably obtained from the Mound Laboratory and was used to analyze the effectiveness of various 

Landauer Neutrak TLDs by the Pinellas health physics department (Burkhart 1987a). It was 

determined that the Landauer dosimeter responded to only about 67% of the dose equivalent for the 

RTG PuO2 heat source spectrum. 
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 26 of 59

Figure 6-2. RTG plutonium source neutron energy spectrum –

PuO2 microspheres (Greene 1984b).

6.3 ADJUSTMENTS TO RECORDED DOSE

Adjustments to Pinellas Plant recorded doses are necessary to arrive at an accurate or favorable-toclaimant dose, considering the uncertainty associated primarily with the complex workplace radiation 

fields and exposure geometries. A key item for dose reconstructors to understand about the GEND 

radiation protection program is that for an individual lifetime dose [denoted as the Neutron Devices 

Department (NDD) lifetime dose in GEND files], the internal exposure from tritium was combined with 

the external exposure from neutrons, gamma rays, and X-rays. To reconstruct a claimant’s lifetime 

dose properly, the dose reconstructor should separate the tritium dose from the rest of the external 

dose. 

The recorded Pinellas doses show that there is not a consistent relationship between recorded 

neutron and photon doses for work performed in the neutron generator areas at the Plant. This lack 

of a true neutron-to-photon dose ratio can be attributed to the nature of the Pinellas processes, during 

which neutron generator testing occurred in open rooms that, combined with the short period of the 
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 27 of 59

neutron pulse, relatively open test structure, virtually no photons from the neutron generator, and 

relative low number of neutrons per test pulse (no significant quantities of activation products), would 

result in a corresponding photon dose. This result is supported in the individual dose records, which 

indicate that the timing of Pinellas personnel neutron and photon exposures varied greatly not only on 

a yearly but also on a monthly basis (a recorded value for 1 month and no recorded doses for the next 

several months). Thus, the assignment of a neutron-to-photon dose ratio to adjust for a missed 

neutron dose is not valid for the Pinellas Plant for neutron generators.

However, for RTG PuO2 heat sources an approximate 3:1 neutron-to-photon ratio was measured 

based on neutron and photon exposure rates from the processing of the RTG PuO2 heat sources as 

measured in 1987 (GE 1987b). Exposure rates measured at another time were 0.37 mR/hr neutron 

and 0.18 mR/hr gamma at 24 in. Actual dose equivalent received from removal and processing of 15 

units for one worker (apparently the typical workload was 50 generators per month for perhaps three 

personnel) amounted to a 20.9-mrem neutron and 7.1-mrem gamma dose equivalents or 

approximately a 3:1 neutron-to-photon ratio (GE 1983). Because the RTG radioisotopic ratios and 

activities remained unchanged and the workstations and processes consistent, this ratio can also be 

considered a constant for RTG work conducted from 1975 to 1990. 

6.3.1 Photon Dose Adjustments

With the exception of the Mound TLDs, no adjustments to photon doses are necessary for dose 

reconstruction.

6.3.1.1 Signal Fading Adjustments

Prior to July 1981, the Mound TLDs were subject to signal fading, which could amount to 25 to 30% of 

the stored signal fading during a period of 4 weeks (Crain 1981a). Beginning with the third quarter of 

1981, the Mound TLDs were annealed before being packed into the dosimeter holders/badges to 

reduce the signal fading problem to a negligible level (Crain 1981a,b). As a result of this signal 

fading, the doses reported for the affected dosimeters were as little as 70% of the actual dose. To 

compensate for the signal fading, the Mound TLD results reported during the period from October 

1979 through June 1981 should be multiplied by a signal fading correction factor of 1.43 (1/0.70). 

After June 1981, there was no significant signal fading for those dosimeters, and no corrections for 

signal fading are necessary.

6.3.2 Neutron Dose Adjustments

The following subsections provide the bases for the various neutron dose adjustments. The 

necessary neutron dose adjustments are summarized in Tables 6-7 and 6-8 below. 

6.3.2.1 Radiation Weighting Factor Adjustments

The Pinellas Plant used a relative biological effectiveness (RBE) weighting factor of 10.0 when 

calculating the effective dose for 14-MeV neutrons (Holliday undated; MMSC 1994b, p. 249), which is 

equivalent to the ICRP Publication 60 neutron weighting factor for neutron energies from 2 to 14 MeV 

(ICRP 1991). Because the weighting factor value used by the Pinellas Plant is higher than other 

values that could have been used (such as those in NCRP 1971), the ICRP Publication 60 correction 

factor is unity for neutron exposures in the neutron generator areas.

The plutonium in the RTGs emits neutrons with an average energy of 2 MeV and an energy range of 

thermal to 12 MeV (Figure 6-2) (Burkhart 1987a). Based on the average neutron energy being 

2 MeV, 50% of the neutrons were assumed to be in the 2- to 20-MeV neutron energy group and the 

remaining 50% of the neutrons were assumed to be in the neutron energy group of 0.1- to 2-MeV, 
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 28 of 59

which is typically the most favorable-to-claimant neutron energy group. The available GEND 

documentation does not describe any separate processing of the dosimetry when Landauer 

dosimeters were used to monitor exposures from RTG operations. Therefore, it was assumed that an 

RBE of 10 was applied to the neutron doses from RTG operations when Landauer dosimeters were 

used for the RTG operations (i.e., 1975–September 1979 and September 1987–February 1991). 

When the Mound dosimeters were used to monitor exposures from RTG operations (i.e., October 

1979–September 1987), a single RBE value of 7 was applied to the Pinellas Plant doses reported by 

the Mound Laboratory (ORAUT 2004). Because the RBE value used for the Landauer dosimeters is 

uncertain and because the RBE used for the Mound dosimeters is likely more favorable to claimants 

than what was used for the Landauer dosimeters, the radiation weighting factor adjustments for all 

neutron energy groups and all periods of RTG operations should be based on an RBE value of 7. 

A summary of the ICRP Publication 60 correction factors (ICRP 1991) for the Pinellas Plant’s reported 

neutron doses is provided in Table 6-7.

Table 6-7. Neutron radiation energies and ICRP 60 correction factors.

Process 

Type

Neutron

Energy

Range

IREP

Energy

Group

Default Dose

Fractiona

(%)

ICRP 60/RBE

Ratiob

ICRP 60 

Correction 

Factorc

neutron 

generator 

operations

≤14 MeV 2-20 MeV 100% 10/10 1.00

RTG 

operations

0-2 MeV

2-12 MeV

0.1-2 MeV

2-20 MeV

50%

50%

20/7

10/7

1.43

0.71

a. Because no neutron dose values for the various neutron energies were available at the time of this 

assessment, the default dose fractions for the Pinellas Plant are actually just based on the expected neutron 

energy distributions.

b. The ICRP 60/RBE ratio is the appropriate radiation weighting factor in ICRP 60 (ICRP 1990) divided by the 

historical RBE value that was applied to the reported neutron doses.

c. The ICRP 60 correction factor is the ICRP 60/RBE ratio multiplied by the appropriate fraction/percentage for 

 the neutron energy group. 

6.3.2.2 Track Fading Adjustments

The NTA track film that was used at the Pinellas Plant from the start of operations in 1957 through 

June 1978 was susceptible to track fading. Track fading occurs between the time the tracks in the 

dosimeter were created and when the dosimeter film was analyzed. As this time interval increases, 

the amount of track fading increases.

A study was performed in 1969 to determine the amount of track fading on the NTA track film 

dosimeters that were used at the Pinellas Plant. Track fading was potentially significant because of 

the Plant’s monthly dosimeter exchange frequency. The study determined that an average of 67% of 

the proton-recoil tracks had faded for a monthly dosimeter exchange. In other words, on average only 

33% of the original tracks remained by the time the dosimeters were analyzed. The study 

recommended that a correction factor of 3 (i.e., 1/0.33) be incorporated into the dose calculations 

beginning in January 1970 to account for track fading (Author unknown1969). A 1974 memorandum 

from Landauer indicates that the correction factor of 3 was also being applied to the NTA film

dosimeter results provided by Landauer for the period of July 1974 through May 1978 (Wheeler 

1974).

Because there is no indication that the Pinellas Plant performed any track fading corrections to the 

reported neutron doses prior to 1970, the dose reconstructor should apply a track fading correction 

factor of 3.0 to the neutron doses reported for the years of 1957–1969. No track fading corrections 
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 29 of 59

need to be applied to the neutron doses reported after 1969, because the reported neutron doses 

have already been corrected.

6.3.2.3 Signal Fading Adjustments

Prior to July 1981, the Mound TLD was subject to signal fading, which could amount to 25% to 30% of 

the stored signal fading during a period of 4 weeks (Crain 1981a). Beginning with the third quarter of 

1981, the Mound TLDs were annealed before being packed into the dosimeter holders/badges to 

reduce the signal fading problem to a negligible level (Crain 1981a,b). As a result of this signal 

fading, the doses reported for the affected dosimeters were as little as 70% of the actual dose. To 

compensate for the signal fading, the Mound TLD results reported during the period from October 

1979 through June 1981 should be multiplied by a signal fading correction factor of 1.43 (1/0.70). 

After June 1981, there was no significant signal fading for those dosimeters, and no corrections for 

signal fading are necessary.

6.3.2.4 Neutron Energy Response Adjustments

Because of the higher neutron energies associated with the neutron generator production activities, 

the reported neutron doses for workers in the neutron generator portions of the Pinellas Plant do not 

need to be adjusted for the dosimeters’ poor response to lower neutron energies. However, reported 

neutron doses for the Plant’s RTG workers need to be adjusted because of the limitations of some of 

the dosimeter responses to the lower energy neutrons that were encountered in the RTG areas. 

NTA film has a neutron energy response threshold of 0.5 MeV. Because some of the neutrons 

associated with the RTG PuO2 heat sources had neutron energies below 0.5 MeV, the reported 

neutron doses from the start of Pinellas operation through June 1978 require a correction for poor 

energy response. Based on the neutron energy distribution information for the RTG PuO2 heat 

sources (Burkhart 1987a), 6.4% of the neutrons were in the 0–1 MeV energy range and were 

assumed to not have been detected by the NTA film dosimeters. Therefore, a neutron energy 

response correction factor of 1.07 [1/(1-0.064)] should be applied to all NTA film dosimeter results to 

account for any under-reported neutron doses due to the NTS film dosimeter’s energy response. 

Because the Mound dosimeters utilized TLD chips that responded well to the neutron energies 

encountered by RTG workers, no adjustment for the neutron energy response of the Mound 

dosimeters is necessary (ORAUT 2004).

The GEND health physics group determined that the Neutrak ER portions of the Landauer Type R/I 

neutron dosimeter recorded only 67% of total neutron dose because of the dosimeter’s poor energy 

response to low-energy neutrons from the RTG PuO2 heat sources (Burkhart 1987a). In addition, 

there is no indication that the reported neutron doses for the RTG workers have been adjusted by the 

Pinellas Plant. Therefore, a neutron energy response correction factor of 1.49 (1/(0.67) would 

account for the under-reported neutron doses from the RTG PuO2 heat sources. 
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 30 of 59

Table 6-8. Summary of neutron dosimeter correction factors.a

Applicable period

Type of correction factor

Total

adjustment

Track

fading

Signal

fading

Energy

response

Neutron generator operations (1957–1997)

1957–1969 3.00 N/A N/A 3.00

1970–1997 N/A N/A N/A 1.00

RTG operations (1975–1991)

1975–Sep 1979 N/A N/A 1.07 1.07

Oct 1979–Jun 1981 N/A 1.43 N/A 1.43

Jul 1981–Sep 1987 N/A N/A N/A 1.00

Oct 1987–Feb 1991 N/A N/A 1.49 1.49

a. This summary includes all of the neutron dosimeter correction factors except the 

ICRP Publication 60 correction factors, which are already summarized in 

Table 6-7.

6.3.3 Electron Dose Adjustments

Electron doses were monitored but not routinely recorded. The primary source of electron exposures 

was from the use of two85Kr leak detection systems (Radiflo® and TracerFlo systems) in Area 109 

from about 1963 through 1994. 

Beta dose monitoring for 85Kr started before the DOELAP standard release, which used a calibration 

factor from a 90Sr/90Y source that tended to underestimate the dose from 85Kr exposures. To 

compensate for the lower energy of 85Kr in relation to that of 90Sr/90Y, a correction factor might have 

been used for the Pinellas Plant based on the more similar 204Tl energy spectrum. Because it is not 

clear from the Plant records whether the 204Tl energy calibration or equivalent was requested by 

Pinellas of Landauer or other vendors prior to 1986, a correction should be performed on the reported 

non-penetrating or electron doses prior to 1986 to ensure that the workers’ doses are not 

underestimated. To compensate for the energy spectrum differences the reported nonpenetrating or 

electron doses should be multiplied by a correction factor of 3.5 for the years prior to 1986 (Poliziani

1985). From 1986 onwards, when DOELAP and National Voluntary Laboratory Accreditation 

Program standards included 204Tl calibration criteria for 85Kr exposures, it is not necessary to apply a 

correction factor. 

6.4 MISSED DOSE

Missed external dose is the unrecorded or unmeasured external dose that is the result of either

relatively high detection limits, short monitoring periods, high dosimeter exchange frequencies, or a 

combination of these three factors (NIOSH 2007). Missed doses are applicable to the Pinellas Plant 

workers that were monitored for external dose and had one or more reported dosimeter readings that 

were less than half the LOD for the dosimeter. Missed dose is primarily estimated based on 

dosimeter results, where the number of zero or < LOD/2 values for a given year is multiplied by the 

LOD/2 value for the dosimeters used during that year (NIOSH 2007). 

At the Pinellas Plant, electron, photon, and neutron doses were possible. However, electron doses, 

including missed electron doses, were unlikely, as discussed in Section 6.2.4.1. Based on the 

available information for the Pinellas Plant, chronic external exposures to electrons did not occur and 

acute external exposures to electrons were unlikely. Therefore, missed electron doses do not 

normally need to be assessed for Pinellas Plant workers. 

Tables 6-9, 6-10, and 6-11 summarize the dosimetry parameters to be used for calculating missed 

doses for Pinellas Plant workers.
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 31 of 59

Table 6-9. Whole-body beta-gamma dosimetry missed doses.a

Dosimeter type Period of use LOD (rem)

Exchange 

frequency

Maximum annual missed

dose (rem)b

Whole-body dosimeters for neutron generator operations (1957–1997)

Photographic film 1957–Jun 1974 0.040 photonsc

0.040 electronsc Monthly 0.240 photons0.240 electrons

Landauer Type G/Pd Jul 1974–Apr 1990 0.010 photons

0.040 electrons

Monthly

(1974–1989)

Quarterly

(after 1989)

0.060 photons (monthly)

0.240 electrons (monthly)

0.020 photons (quarterly)

0.080 electrons (quarterly)

Landauer Type Z/Fe May 1990–1997 0.010 photons

0.040 electrons

Quarterly 0.020 photons

0.080 electrons

Whole-body dosimeters for RTG operations (1975–1991)

Landauer Type G/Pd 1975–Sep 1979 0.010 photons

0.040 electrons

Monthly 0.060 photons

0.240 electrons

Mound TLD Oct 1979–Sep 1987f 0.020 photonsg Monthly 0.120 photons

Landauer Type R/I Oct 1987f–Feb 1991 0.010 photons

0.040 electrons

Monthly

(1987–1989)

Quarterly

(after 1989)

0.060 photons (monthly)

0.240 electrons (monthly)

0.020 photons (quarterly)

0.080 electrons (quarterly)

a. Some of the beta-gamma dosimeters listed in this table included a component for monitoring neutron exposures; 

however, only the beta-gamma components of those dosimeters are addressed in this table.

b. The maximum annual missed dose assignments are calculated by multiplying half of the dosimeter’s LOD value by the 

maximum number of dosimeters exchanged in a given year (NIOSH 2007).

c. Estimated LODs for commonly used photon dosimetry (ORAUT 2006b).

d. The Landauer Type P dosimeter was a beta-gamma-neutron dosimeter that likely utilized a Type G dosimeter or 

equivalent for its beta-gamma component.

e. After 1994, the Landauer Type Z dosimeter was renamed the Type F dosimeter.

f. Dual Mound and Landauer Type R neutron dosimeters were actually used to monitor RTG workers during the month of 

September in 1987 (Burkhart 1987b). Because it is uncertain which dosimeters were used for the reported doses, the 

dosimeter with the most favorable-to-claimant LOD values was used and the period of use date has been adjusted to 

reflect that.

g. The LODs for the Mound dosimeters are based on the site-specific information provided for the Mound Site (ORAUT 

2004).
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 32 of 59

Table 6-10. Whole-body neutron dosimetry missed doses.

a

Dosimeter

type Period of use LOD (rem)

Exchange

frequency

Maximum annual

missed dose (rem)b

Whole-body dosimeters for neutron generator operations (1957–1997)

NTA film 1957–Jun 1974 0.050 neutronsc Monthly 0.300 neutrons

Landauer Type Pd Jul 1974–May 1978 0.040 neutronse Monthly 0.240 neutrons

Landauer Type E Jun 1978–1997 0.020 neutrons

e Monthly 

(1974–1989)

Quarterly 

(after 1989)

0.120 neutrons (monthly)

0.040 neutrons (quarterly)

Whole-body dosimeters for RTG operations (1975–1991)

Landauer Type Pd 1975–Sep 1979 0.040 neutronse Monthly 0.240 neutrons

Mound TLD Oct 1979–Aug 1987f 0.010 neutronsg Monthly 0.060 neutrons

Landauer Type R/I Sep 1987f–Feb 1991 0.020 neutronse Monthly

(1987–1989)

Quarterly

(after 1989)

0.120 neutrons (monthly)

0.040 neutrons (quarterly)

a. Some of the neutron dosimeters listed in this table included a component for monitoring beta-gamma exposures; 

however, only the neutron components of those dosimeters are addressed in this table.

b. The maximum annual missed dose assignments are calculated by multiplying half of the dosimeter’s LOD value by the 

maximum number of dosimeters exchanged in a given year (NIOSH 2007).

c. NTA film LOD value based on Wilson et al. (1990).

d. The Landauer Type P dosimeter was a beta-gamma-neutron dosimeter that utilized NTA film for its neutron component.

e. The LOD for this Landauer NTA film dosimeter is based on specifications obtained from Passmore (2004).

f. Dual Mound and Landauer Type R neutron dosimeters were actually used to monitor RTG workers during September 

1987 (Burkhart 1987b). Because it is uncertain which dosimeters were used for the reported doses, the dosimeter with 

the most favorable-to-claimant LOD values was used and the period of use date has been adjusted to reflect that.

g. The LODs for the Mound dosimeters are based on the site-specific information provided for the Mound Site (ORAUT 

2004). 
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 33 of 59

Table 6-11. Extremity beta-gamma dosimetry missed doses. 

Dosimeter type Period of use LOD (rem)

Exchange 

frequency

Maximum annual

missed dose (rem)a

Wrist dosimeters

Photographic film 1957–Jun 1974 0.040 photonsb

0.040 electronsb

Monthly 0.240 photons

0.240 electrons

Landauer Type G Jul 1974–1990 0.010 photons 

0.040 electrons

Monthly

(1974–1989)

Quarterly

(after 1989)

0.060 photons (monthly)

0.240 electrons (monthly)

0.020 photons (quarterly)

0.080 electrons (quarterly)

Landauer Type K 1991–1997 0.010 photons

0.040 electrons

Quarterly 0.020 photons

0.080 electrons

Finger ring dosimeters

Photographic film 1957–Jun 1974 0.040 photonsb

0.040 electronsb

Monthly 0.240 photons

0.240 electrons

Unknownc Jul 1974–1982

Landauer Type U About 1983d–1997 0.030 photons

0.040 electrons

Monthly

(1974–1989)

Quarterly

(after 1989)

0.180 photons (monthly)

0.240 electrons (monthly)

0.030 photons (quarterly)

0.080 electrons (quarterly)

a. The maximum annual missed dose assignments are calculated by multiplying half of the dosimeter’s LOD value by the 

maximum number of dosimeters exchanged in a given year (NIOSH 2007).

b. Estimated LODs for commonly used photon dosimetry (ORAUT 2006b).

c. Because the type of ring dosimetry used during this period was unknown at the time of this analysis, the assessment of 

any ring dosimeter results during this period will need to be handled on a case-by-case basis.

d. This is an estimated date because this analysis found no documentation that shows the start of use for the Landauer 

Type U dosimeter. 

6.5 UNMONITORED DOSE

6.5.1 Unmonitored Workers

The majority of the work performed at the Pinellas Plant did not involve exposures to external sources 

of radiation, which explains why a significant number of the workers were not monitored for external 

dose. Based on the review of the available dosimetry data, employees with any significant potential 

for external dose exposure appear to have been routinely monitored, as evidenced by the large 

number of monitored individuals that routinely had doses below the reporting levels. Therefore, it is 

reasonable to assume that unmonitored workers received less dose than monitored workers at the 

Pinellas Plant.

For the periods in which any Pinellas Plant worker was not monitored for external dose, an annual 

unmonitored external dose assignment of 100 mrem should be assigned. For unmonitored periods 

that are less than 1 year in duration, the unmonitored dose assignment should be prorated, unless 

doses are being overestimated. For Plant workers who were likely exposed only to the onsite ambient 

levels of radiation, this more favorable-to-claimant unmonitored dose is assigned in lieu of the onsite

ambient doses provided in the site’s technical basis document on occupational environmental dose 

(ORAUT 2011c), for the reasons provided in the basis for the unmonitored external dose assignment. 

The basis for this unmonitored external dose assignment is in Attachment B of this document. In 

addition, the unmonitored external dose assignment should be assigned only as 100% 30- to 250-keV 

photons, for the reasons provided in Attachment B.
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 34 of 59

6.5.2 Workers Whose Dosimetry Records Are Missing

DOE has not been able to find the dosimetry records for a significant number of the Pinellas Plant 

workers. This situation was identified by comparing the DOE responses found in the claim files with 

the dosimetry records contained in ORAUT’s Site Research Database (SRDB). For a number of 

claims, DOE has indicated that no records were found, whereas NIOSH has located full or partial sets 

of dosimetry records for those claims in its SRBD. As of the date of this document, DOE has not 

acknowledged that any dosimetry records for Pinellas Plant workers have been lost. 

Because a monitored Pinellas Plant worker had a higher potential to receive external dose than an 

unmonitored worker, special considerations need to be made for workers whose dosimetry records 

are missing. 

• In the instances where sufficient records are found in the SRDB, which are identified as 

Personnel Exposure files in the claim files, no special considerations need to be made and the 

other sections of this document can be used to assess the worker’s external doses. 

• In instances where there is only an indication that a worker was monitored for dose (e.g. 

records of annual whole-body doses found for some years; see note below), the worker’s job 

description should be carefully evaluated to determine if the worker was likely monitored for 

neutron doses and/or photon doses. The appropriate annual missed doses should be 

assigned along with the annual unmonitored external dose assignment prescribed in Section 

6.5.1, because missed doses were not accounted for in the unmonitored dose assignment for 

reasons indicated in Attachment B. For overestimating doses, the annual missed doses can 

be assigned based on the maximum zeros approach (the maximum dosimeters exchanged 

per year approach). For a reasonable, yet likely favorable-to-claimant estimate of a worker’s 

missed doses, the annual missed doses should be based on the maximum potential number of 

dosimeter exchanges within a year minus 1 to account for the unmonitored dose assignment 

as being a single positive dosimeter reading.

Note: When records consisting of only annual whole-body doses are found for some 

years, the reported doses could consist of external photon doses, external neutron 

doses, internal tritium doses, or any combination of these three types of dose. 

However, the available information for the Pinellas Plant indicates that neutron dose 

monitoring was always performed in conjunction with photon dose monitoring. 

Because the reported whole-body doses could consist of only internal tritium doses, 

the dose reconstructor needs to carefully evaluate the employment information for the 

worker to determine what type of dose monitoring was most likely performed. In the 

event that the type of dose monitoring cannot be determined, the default assumption 

is that external photon and internal tritium monitoring were performed, because these 

are the two most common types of monitoring that were performed at the Plant.

6.6 UNCERTAINTY

When a reported external dose is based on a single measurement, the uncertainty associated with 

that individual measurement needs to be accounted for in the assigned dose. In contrast to individual 

dose measurements, the uncertainty associated with external doses based on multiple measurements 

for the same period has already been accounted for in the dose assignment. An example of this is 

unmonitored dose assignments, which are typically based on either average or upper bound annual 

doses for a population of worker doses. As a result, this section only addresses how the uncertainties 

for individual external dose measurements should be assessed.
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 35 of 59

6.6.1 Beta-Gamma Dosimeter Uncertainty

For film dosimeters, the LODs that are quoted in the literature range from about 30 to 50 mrem for 

electron and photon irradiation (Morgan 1961). These are not the expected uncertainties at larger 

electron and photon dose readings. For example, it was possible to read a photon dose of 100 mrem 

to within ±15 mrem (±15%) if the exposure involved photons with energies between several hundred 

keV and several MeV (Morgan 1961). If the exposure involved photons with energies less than 

several hundred keV, the uncertainty was at least twice that for the more energetic photons (Morgan 

1961). Therefore, the standard error in the recorded film dosimeter doses from photons of any energy 

is estimated to be ±30%. The standard error for the recorded dose from electron irradiation was 

essentially the same as that for photon irradiation, but when an unknown mixture of electron and 

photon irradiation was involved, the standard error for the dose from beta irradiation was somewhat 

larger than ±30% (Morgan 1961). Therefore, the standard error in the recorded film dosimeter doses 

from >15 keV electrons is estimated to be ±30% for a known mixture of photon and electron 

irradiation, and higher if the mixture is unknown.

For TLDs, the uncertainty is generally lower than the uncertainty for film dosimeter results; however, 

the uncertainty is still somewhat dependent on the measured dose (NIOSH 2007). Based on that 

observation, the uncertainty associated with the Pinellas Plant’s recorded electron and photon doses 

from TLDs will be assumed to be ±30%, which is potentially favorable to claimants for some reported 

dose values. 

6.6.2 Neutron Dosimeter Uncertainty

The NTA film technology used to measure neutron doses at the Pinellas Plant was similar to the 

technology used at other AEC/DOE facilities. Based on a review of the available information for such

facilities, a reasonable uncertainty for neutron dose measurements performed using NTA film 

dosimeters is ±50%. 

For the Landauer Type E neutron dosimeters, which were a polycarbonate (Lexan®) dosimeter, the 

sensitivity to fast neutron radiation was reported in 1978 as 30 ± 15 mrem (±50%) (Holliday 1978). It 

was also verified that the uncertainties of the dose determinations decrease as exposures increase 

(Holliday 1978). Therefore, a reasonable uncertainty for neutron dose measurements performed 

using the Landauer Type E dosimeters is ±50%. 

For the Mound TLDs, an uncertainty of ±30% should be applied, based on the site-specific information 

for the Mound Site (ORAUT 2004). 

For the Landauer Type R/I neutron dosimeters, an uncertainty of ±30% should be considered 

reasonable and likely favorable to claimants, based on a review of the available information for other 

AEC/DOE facilities.

6.7 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 36 of 59

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Neutron Devices Department), R. S. Landauer, Jr. & Co., Glenwood, Illinois, October 10. 

[SRDB Ref ID: 15848, pages 1534–1535]

Wilson, R. H., J. J. Fix, W. V. Baumgartner, and L. L. Nichols, 1990, Description and Evaluation of the 

Hanford Personnel Dosimeter Program from 1944 through 1989, PNL-7447, Battelle Memorial 

Institute, Pacific Northwest Laboratory, Richland, Washington, September. [SRDB Ref ID: 

4793]

Yoder, R. C., 2005, “Landauer Film Dosimeter Information,” e-mail to Leo Faust from R. Craig Yoder, 

Landauer, Inc., Glenwood, Illinois, December 8. [SRDB Ref ID: 19707] 
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 43 of 59

GLOSSARY

absorbed dose, D

Amount of energy (ergs or joules) deposited in a substance by ionizing radiation per unit mass 

(grams or kilograms) of the substance and measured in units of rads or grays. See dose.

accreditation

In relation to this document, recognition that a dosimeter system has passed the performance 

criteria of the DOE Laboratory Accreditation Program (DOELAP) standard (DOE 1986) in 

specified irradiation categories.

accuracy

The characteristics of an analysis or determination that ensures that both the bias and 

precision of the resultant quantity will remain within the specified limits.

albedo dosimeter

Thermoluminescent dosimeter that measures the thermal, intermediate, and fast neutrons 

scattered and moderated by the body or a phantom from an incident fast neutron flux.

alpha radiation 

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2. 

backscatter

Reflection or refraction of radiation at angles over 90 degrees from its original direction.

beta particle

See beta radiation.

beta radiation 

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

curie (Ci) 

Traditional unit of radioactivity equal to 37 billion (3.7 x 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

densitometer

Instrument that uses a photoelectric cell to measure the transition of light through developed 

X-ray film to determine the optical density.

DOE Laboratory Accreditation Program (DOELAP) 

Program for accreditation by DOE of DOE site personnel dosimetry and radiobioassay 

programs based on performance testing and the evaluation of associated quality assurance, 

records, and calibration programs. 

dose 

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sievert; other types of dose are in 

units of roentgens, rads, reps, or grays.
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 44 of 59

dose equivalent (H)

In units of rem or sievert, product of absorbed dose in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose.

dose equivalent index

Historical measure for neutron source calibration defined by the International Commission on 

Radiation Units and Measurements as the sum of the maximum dose equivalents delivered 

within a sphere at any depth for the respective neutron energies even though the maximum 

dose occurred at different depths and discounting the outer 0.07-millimeter-thick shell. Also 

called unrestricted dose equivalent index.

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See albedo dosimeter, film dosimeter, neutron film dosimeter, and 

thermoluminescent dosimeter.

dosimetry system

System for assessment of received radiation dose. This includes the fabrication, assignment, 

and processing of external dosimeters, and/or the collection and analysis of bioassay samples, 

and the interpretation and documentation of the results. 

electron radiation 

See beta radiation.

error

Difference between the correct, true, or conventionally accepted value and the measured or 

estimated value. Sometimes used to mean estimated uncertainty. See accuracy and 

uncertainty. 

exchange period (frequency)

Period (weekly, biweekly, monthly, etc.) for routine exchange of dosimeters. Also called 

exchange frequency.

exposure

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

extremities 

The portion of the arm from and including the elbow through the fingertips and the portion of 

the leg from and including the knee and patella through the toes. 

fast neutron

Neutron with energy equal to or greater than 10 kiloelectron-volts. This type of neutron causes 

fission in some isotopes (e.g., 238U, 239Pu). See intermediate neutron and slow neutron.

favorable to claimants

In relation to dose reconstruction for probability of causation analysis, having the property of 

ensuring that there is no underestimation of potential dose, which often means the assumption 

of a value that indicates a higher dose than is likely to have actually occurred in the absence of 

more accurate information. See probability of causation.
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film

In the context of external dosimetry, radiation-sensitive photographic film in a light-tight 

wrapping. See film dosimeter.

film dosimeter

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge.

filter

Material used in a dosimeter to adjust radiation response to provide an improved tissue 

equivalent or dose response.

gamma radiation 

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are identical 

to X-ray photons of high energy; the difference is that X-rays do not originate in the nucleus. 

intermediate neutron 

Neutron with energy between 0.5 electron-volts and 10 kiloelectron-volts. See fast neutron

and slow neutron.

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons. See 

alpha radiation, beta radiation, gamma radiation, neutron radiation, photon radiation, and X-ray 

radiation.

isotope 

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties.

kilo-electron volt (keV)

Unit of particle energy equal to 1,000 (1 × 103

) electron-volts.

luminescence

Emission of light from a material as a result of some excitation. See thermoluminescence.

limit of detection (LOD)

Minimum level at which a particular device can detect and quantify exposure or radiation. Also 

called lower limit of detection and detection limit or level.

megaelectron-volt (MeV) 

Unit of particle energy equal to 1 million (1 × 106

) electron-volts. 
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monitoring 

Periodic or continuous determination of the presence or amount of ionizing radiation or 

radioactive contamination in air, surface water, groundwater, soil, sediment, equipment 

surfaces, or personnel (for example, bioassay or alpha scans). In relation to personnel, 

monitoring includes internal and external dosimetry including interpretation of the 

measurements. 

neutron (n) 

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

neutron film dosimeter 

Film dosimeter with a nuclear track emulsion, type A, film packet.

neutron radiation 

Radiation that consists of free neutrons unattached to other subatomic particles emitted from a 

decaying radionuclide. Neutron radiation can cause further fission in fissionable material such 

as the chain reactions in nuclear reactors, and nonradioactive nuclides can become 

radioactive by absorbing free neutrons. See neutron.

personal dose equivalent [Hp(d)] 

Dose equivalent in units of rem or sievert in soft tissue below a specified point on the body at 

an appropriate depth d. The depths selected for personal dosimetry are 0.07 millimeters 

(7 milligrams per square centimeter) and 10 millimeters (1,000 milligrams per square 

centimeter), respectively, for the skin (shallow) and whole-body (deep) doses. These are 

noted as Hp(0.07) and Hp(10), respectively. The International Commission on Radiological 

Measurement and Units recommended Hp(d) in 1993 as dose quantity for radiological 

protection. 

photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz. 

photon radiation 

Electromagnetic radiation that consists of quanta of energy (photons) from radiofrequency 

waves to gamma rays. 

probability of causation (POC)

For purposes of dose reconstruction for the Energy Employees Occupational Illness 

Compensation Program Act, the percent likelihood, at the 99th percentile, that a worker 

incurred a particular cancer from occupational exposure to radiation.

quality factor, Q

Principal modifying factor (which depends on the collision stopping power for charged 

particles) that is employed to derive dose equivalent from absorbed dose. The quality factor 

multiplied by the absorbed dose yields the dose equivalent. See dose and relative biological 

effectiveness.

rad

Traditional unit for expressing absorbed radiation dose, which is the amount of energy from 

any type of ionizing radiation deposited in any medium. A dose of 1 rad is equivalent to the 

absorption of 100 ergs per gram (0.01 joules per kilogram) of absorbing tissue. The rad has
Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 47 of 59

been replaced by the gray in the International System of Units (100 rads = 1 gray). The word 

derives from radiation absorbed dose. 

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer.

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei.

radioisotopically-powered thermoelectric generator (RTG)

Generator that obtains its power from passive (natural) radioactive decay using thermocouples 

to convert the heat of decay into electricity.

random error

When a given measurement is repeated and the values do not agree exactly. The causes of 

the disagreement between the values must also be the causes of their differences from the 

true value. See systematic error. 

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

roentgen

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

or negative charge equal to 2.58 × 10-4 coulombs per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0°C and standard atmospheric pressure. An 

exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft tissue for 

higher energy photons (generally greater than 100 kiloelectron-volts). 

scattering

Change in direction of radiation by refraction or reflection, often accompanied by a decrease in 

radiation due to absorption by the refracting or reflecting material.

shallow dose equivalent

Dose equivalent in units of rem or sievert at a depth of 0.07 millimeters (7 milligrams per 

square centimeter) in tissue equal to the sum of the penetrating and nonpenetrating doses.

shielding

Material or obstruction that absorbs ionizing radiation and tends to protect personnel or 

materials from its effects.

skin dose

See shallow dose equivalent.

systematic error

When a given measurement is repeated and the values differ from the true value by the same 

amount. See random error.
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thermal neutron 

Neutron in thermal equilibrium with its surroundings having an average energy of 

0.025 electron-volts. 

thermoluminescence 

Property that causes a material to emit light as a result of heat.

thermoluminescent dosimeter (TLD) 

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated by radiation, release the stored energy as light. The measurement of this 

light provides a measurement of absorbed dose. 

tissue equivalent

Substance with response to radiation equivalent to tissue. A tissue-equivalent response is an 

important consideration in the design and fabrication of radiation measuring instruments and 

dosimeters.

U.S. Atomic Energy Commission (AEC) 

Federal agency created in 1946 to assume the responsibilities of the Manhattan Engineer 

District (nuclear weapons) and to manage the development, use, and control of nuclear energy 

for military and civilian applications. The U.S. Energy Research and Development 

Administration and the U.S. Nuclear Regulatory Commission assumed separate duties from 

the AEC in 1974. The U.S. Department of Energy succeeded the U.S. Energy Research and 

Development Administration in 1979.

uncertainty

Standard deviation of the mean of a set of measurements. The standard error reduces to the 

standard deviation of the measurement when there is only one determination. See accuracy

and error. Also called standard error. 

whole-body dose

Dose to the entire body excluding the contents of the gastrointestinal tract, urinary bladder, 

and gall bladder and commonly defined as the absorbed dose at a tissue depth of 

10 millimeters (1,000 milligrams per square centimeter). Also called penetrating dose. 

See dose.

X-ray

See X-ray radiation.

X-ray radiation 

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom. 
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ATTACHMENT A

EXAMPLES OF PINELLAS PLANT EXPOSURE RECORDS

Page 1 of 5
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ATTACHMENT A

EXAMPLES OF PINELLAS PLANT EXPOSURE RECORDS

Page 2 of 5
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ATTACHMENT A

EXAMPLES OF PINELLAS PLANT EXPOSURE RECORDS

Page 3 of 5
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ATTACHMENT A

EXAMPLES OF PINELLAS PLANT EXPOSURE RECORDS

Page 4 of 5
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ATTACHMENT A

EXAMPLES OF PINELLAS PLANT EXPOSURE RECORDS

Page 5 of 5
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Background

Based on an evaluation of the available dosimetry records for the Pinellas Plant, personnel who had 

any potential to receive significant external dose were likely monitored. In addition, the majority of the 

work performed at the Pinellas Plant was nonradiological, which explains why a significant number of 

the employees were not monitored. However, there was the potential to receive incidental external 

doses that were small (i.e., < 20 mrem/yr) but were much more significant than the highest onsite

ambient doses for the plant [i.e., >> 3.66 × 10-13 mrem/yr, from the Plant’s environmental dose TBD

(ORAUT 2011c)]. One of the most common types of incidental radiation exposures was attributable 

to Plant tours, which included visits to the neutron generator testing areas. Based on information 

provided in the telephone interviews, personnel on these tours were allowed to witness the neutron 

generators being tested without being provided with dosimetry. Based on information that has been 

provided in some Pinellas Plant records, doses at the 3-ft exclusion boundary for the neutron 

generator tests were 9.9 mrem per test. Assuming that personnel on a Plant tour would not have 

likely stayed in the neutron generator testing area for more than two tests, these exposed personnel 

might have received an incidental dose of almost 20 mrem. 

Because the incidental doses at the Pinellas Plant were likely much larger than the onsite ambient 

doses, it is not appropriate to assign onsite ambient doses to all unmonitored workers unless it can be 

established that there was no potential for incidental dose due to an employee’s job. Because a 

determination that incidental doses were not received can rarely be defended, unmonitored external 

doses should be assigned for all workers in lieu of the onsite ambient doses for the Plant. 

Basis for Upper-95th Percentile Dose of 100 mrem/yr for Unmonitored Workers

A review of the available dosimetry data for the Pinellas Plant indicates that on average 95% of the 

monitored workers at the plant received annual doses ≤ 100 mrem. This determination is based on 

an evaluation of whole-body dose information that was available in the following documents: GE 

1956–1980, 1957–1990, 1975; Greene 1985b, 1986; Holliday 1977, 1979, 1981, 1983b; Jech 1963; 

LMSC 1996; Schumacher 1982, 1983, 1984; and Weaver 1994b.

Whole-body dose information was used for the evaluation because it was the only form of data that 

was consistently available throughout the years the Pinellas Plant operated. Unfortunately, the 

Plant’s reported whole-body doses consist of external photon, external neutron, and internal tritium 

doses, and the contribution attributable to each of these components cannot be determined for most 

years. In addition, the doses evaluated represent population doses that do not account for potential 

missed doses. Table B-1 provides summarized results of the dosimetry data that were evaluated. 

The average percentage of annual whole-body doses below 100 mrem was 94.7% (95% when 

rounded to the nearest percent), for the years that were evaluated. However, for a select number of 

years, more detailed dose information was provided, and it was determined that a significant number 

of the annual whole-body doses were either entirely or mostly attributable to tritium dose. This is 

consistent with the types of operations performed at the Pinellas Plant. Therefore, the upper-95thpercentile external dose for monitored workers is likely to be less than 100 mrem/yr.

ATTACHMENT B

BASIS FOR THE UNMONITORED EXTERNAL DOSE ASSIGNMENT

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Table B-1. Evaluation of Pinellas Plant doses available in the SRDB.

Year

Total number of 

annual doses 

checked

Number of

annual doses 

≥ 100 mrem

Percentage of

annual doses 

below 100 mrem

Number of

annual doses 

≤ 20 mrem

Percentage of

annual doses ≤ 

20 mrem

1957 71 12 83.1 38 53.5

1958 142 4 97.2 102 71.8

1959 205 43 79.0 114 55.6

1960 232 40 82.8 147 63.4

1961 265 39 85.3 168 63.4

1962 259 36 86.1 159 61.4

1963 258 20 92.2 197 76.4

1964 292 13 95.5 212 72.6

1965 280 37 86.8 233 83.2

1966 320 17 94.7 277 86.6

1967 351 8 97.7 328 93.4

1968 355 3 99.2 340 95.8

1969 344 9 97.4 326 94.8

1970 293 9 96.9 258 88.1

1971 282 6 97.9 261 92.6

1972 292 7 97.6 267 91.4

1973 280 10 96.4 254 90.7

1974 303 8 97.4 265 87.5

1975 276 2 99.3 246 89.1

1976 255 5 98.0 195 76.5

1977 246 27 89.0 159 64.6

1978 262 20 92.4 176 67.2

1979 313 17 94.6 244 78.0

1980 383 23 94.0 249 65.0

1981 388 18 95.4 324 83.5

1982 418 4 99.0 286 68.4

1983 396 11 97.2 263 66.4

1984 428 13 97.0 296 69.2

1985 395 15 96.2 253 64.1

1986 322 NAa NAa NAa NAa

1987 248 NAa NAa NAa NAa

1988 263 4 98.5 245 93.2

1989 285 5 98.2 270 94.7

1990 284 4 98.6 258 90.8

1991 288 1 99.7 274 95.1

1992 280 0 100.0 275 98.2

1993 245 0 100.0 242 98.8

1994 NA NAa NAa NAa NAa

1995 215 0 100.0 NAa NAa

1996 NAa NAa NAa NAa NAa

1997 NAa NAa NAa NAa NAa

1998 NAa NAa NAa NAa NAa

Averages 94.7 79.6

a. NA – not available.

ATTACHMENT B

BASIS FOR THE UNMONITORED EXTERNAL DOSE ASSIGNMENT

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Upon review of the dosimetry data in the referenced documents listed above and Table 6-1 of this 

document, many discrepancies regarding the numbers of individuals monitored were noted. 

Therefore, it is not recommended that the 95th-percentile doses for individual years be tabulated 

unless a better data set is located. By averaging the data over several years, the potential for 

incurring a significant error for a given year would be minimized. In addition, including the dose 

contribution attributable to tritium helps to compensate for any potential errors and helps to ensure 

that the 95th-percentile dose is favorable to claimants. Because the majority (>79% each year on 

average) of the annual whole-body doses reported for Pinellas Plant workers were ≤ 20 mrem, which 

is an individual dosimeter’s limit of detection for some of the dosimeters used at the Plant, it appears 

that the Plant monitored workers if they had any potential to receive dose. As a result, it is unlikely 

that any unmonitored workers would have received doses approaching the upper-95th-percentile 

doses. Therefore, it is very unlikely that any errors in the 95th-percentile dose calculation attributable 

to these discrepancies would result in an unfavorable to claimant dose assignment for unmonitored 

individuals.

Comparison of Maximum Likelihood Doses for Unmonitored Workers to the Unmonitored Dose 

Assignment

Area dosimeter results with their placement locations identified in relation to the external radiation 

source terms are very limited. However, detailed area dosimeter results were available for the period 

from 1984 through June 1986 in GE (1984–1986). 

The area dosimeter results available for Areas 128, 131, and 183 were evaluated. These areas 

represent photon and neutron exposure areas, which are in the neutron generator operational areas 

of the Plant. Area 131 was the final testing area of the neutron generators. Area 128 was 

represented by dosimeter locations 9000, 9001, 9003, 9004, 9008, and 9019. Area 131 was 

represented by dosimeter locations 9026 and 9038. Area 183 was represented by dosimeter location 

9020. All of these dosimeter locations were reported as being within 24 to 65 in. of the closest 

radiation source. For the period from 1984 through June 1986, only a single photon dosimeter result 

above the dosimeter’s LOD was reported and only two neutron dosimeter results at two different 

locations were reported as being above the dosimeter’s LOD. 

Based on the area dosimeter data that were evaluated, the worst-case annual photon dose is 

20 mrem plus 11 dosimeters worth of missed dose, and the worst-case annual neutron dose is 

40 mrem plus 11 dosimeters worth of missed dose. Because area dosimeter results correspond to a 

8,760-hour year versus a 2,500-hour work-year, the worst-case annual doses needed to be adjusted 

by a maximized occupancy factor of 0.2854 (2,500/8,760). For the photon and neutron dosimeters 

used during this period, no other dose adjustments were needed as indicated in Section 6.3. For 

1984–1986, the LOD/2 values for the applicable photon and neutron dosimeters were 5 mrem and 10 

mrem per dosimeter exchange, respectively. Based on those values, the maximum likely external 

dose to an unmonitored worker would be <64 mrem/yr (21 mrem/yr from photons and 43 mrem/yr 

from neutrons). Because it is unlikely that a Pinellas Plant worker would have spent 100% of their 

time within 65 in. of a radiation source without being monitored, this is likely a significant overestimate 

of an unmonitored worker’s potential external dose. 

The potential doses to workers in the RTG areas of Building 400 are potentially much higher than 

64 mrem/yr. However, because all types of plutonium are considered to be special nuclear material, 

access to the RTG areas was more strictly limited. Because Pinellas Plant personnel having work 

assignments in radiation areas were required to wear dosimeters per General Operating Procedures 

(Greene 1985b; GE 1982), and because access to the RTG areas would have been controlled for 

ATTACHMENT B

BASIS FOR THE UNMONITORED EXTERNAL DOSE ASSIGNMENT

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security purposes, it is much less likely that personnel would have been able to receive significant 

unmonitored doses in the RTG areas. Therefore, the potential doses for unmonitored workers in the 

RTG areas were also likely less than 64 mrem/yr.

Not Using ORAUT-OTIB-0020 Approach for Calculating Coworker Doses

Because an unmonitored worker’s potential to receive external dose at the Pinellas Plant was 

significantly less than the dosimetry program’s ability to detect doses that low, the unmonitored dose 

approach described in ORAUT-OTIB-0020 (ORAUT 2008), which uses only coworker data, can result 

in an unreasonable overestimate of an unmonitored Pinellas Plant worker’s external doses. This 

determination is supported by an analysis of the available area dosimeter results, which have a longer 

effective monitoring period per dosimeter exchange, and the fact that on average >79% of the annual 

whole-body doses reported for Plant workers were ≤ 20 mrem. 

The ORAUT-OTIB-0020 approach requires a favorable-to-claimant LOD/2 approach for accounting for 

the coworker’s potential missed doses (ORAUT 2008). Based on the maximum likelihood doses 

calculated above, the ORAUT-OTIB-0020 approach results in an unreasonably excessive 

unmonitored dose estimate for unmonitored Pinellas Plant workers. Because of this and because the 

95th-percentile dose is already greater than the maximum likelihood dose, the potential missed doses 

were not factored into the unmonitored dose assignment for the Pinellas Plant. Another alternative 

would be to use the maximum likelihood dose of 64 mrem/yr as the unmonitored dose assignment. 

However, a decision was made to use the more favorable-to-claimant unmonitored dose assignment 

of 100 mrem/yr to ensure that the worker’s external doses are not underestimated.

Assigning Unmonitored Doses as Photon Doses Versus Neutron Doses

Because the unmonitored external dose assignment includes both photon and neutron dose 

components and because the proportions of those dose components are unknown and highly 

variable, an evaluation was performed to determine which of the applicable radiation types and 

energy groups to assign the unmonitored doses as. 

The only photon energy distribution that is used for the external doses for Pinellas Plant workers is 

100% 30-250 keV photons, which is also the most favorable-to-claimant photon energy distribution. 

Therefore, only 30-250 keV photons were evaluated for the potential unmonitored dose assignment. 

Based on the information in the main body of this document, there are only two neutron energy 

distributions for the various areas at the Pinellas Plant. For the neutron generators areas at the Plant, 

the neutron energy distribution was 100% 2-20 MeV neutrons. For the RTG areas in Building 400, the 

neutron energy distribution was 50% 0.1-2 MeV and 50% 2-20 MeV neutrons. Given that it was 

unlikely that a worker would have been able to spend any significant time in the presence of the PuO2

sources in the RTG areas without having to wear a dosimeter, the predominant neutron energies that 

unmonitored workers at the Plant were likely exposed to were 2-20 MeV neutrons. Therefore, only 

2-20 MeV neutrons were evaluated for the potential unmonitored dose assignment.

The unmonitored dose assignment of 100 mrem/yr was assigned as photon dose and then neutron 

dose to compare the POC results using each of the IREP cancer models. Because ORAUT-OTIB0005 (ORAUT 2011a) indicates that there could be multiple external organ dose selections for a given 

IREP cancer model, this evaluation was performed using only the external organ dose selection that 

resulted in the highest dose (i.e., the selection with the highest organ DCF) for a given IREP cancer 

model selection. Each IREP cancer model was evaluated using the following hypothetical scenario 

assumptions:

ATTACHMENT B

BASIS FOR THE UNMONITORED EXTERNAL DOSE ASSIGNMENT

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1. The hypothetical worker was a male, with the exception of cancer models that are only 

applicable to females.

2. The worker was 18 years old at the time of the first exposure. Therefore, a year of birth of 

1939 and a first year of employment date of 1957 were selected.

3. The worker was assumed to be employed and unmonitored for 5 years (i.e., during all of 1957 

through 1961).

4. Only an unmonitored dose of 100 mrem was assigned for each year of employment.

5. The date of diagnosis was set at 12 years after the last year of employment (i.e., 1973).

Because the unmonitored dose assignment represents a population dose, no uncertainty correction 

factors were applied to the unmonitored dose assignment. As indicated in the main body of this 

document, no dosimeter correction factors need to be applied to the photon doses. Even though the 

main body of this document indicates that dosimeter corrections factors would be applied to the 

reported neutron doses for most years, no dosimeter correction factors were applied when the 

unmonitored doses were evaluated as neutron doses. The basis for this is that the unmonitored dose 

assignment is essentially an arbitrary value that has been determined to be more favorable to 

claimants than the maximum likelihood dose for the Pinellas Plant, and the evaluation of the 

maximum likelihood dose for the Plant utilized neutron dosimeter data that did not require any 

adjustments, as indicated in Section 6.3. In addition, the ICRP Publication 60 correction factor for 

2-20 MeV neutron doses is 1.00 for the Plant, so this correction factor had no impact on the neutron 

doses being evaluated (ICRP 1991).

The results of this evaluation are summarized in Table B-2 below. With the only exceptions being the 

acute myeloid leukemia, chronic myeloid leukemia, and leukemia (less chronic lymphocytic leukemia) 

IREP cancer models, the assignment of the unmonitored doses as 100% 30-250 keV photons 

provides a more favorable-to-claimant POC. At the time of this document’s preparation, it was 

determined that out of 393 Pinellas Plant claims, which excludes pulled claims, there are only 5 

leukemia claims that are affected by the photon-versus-neutron unmonitored dose issue. However, 

because the unmonitored workers at any site only need to be assigned 50th-percentile coworker 

doses and because the unmonitored external dose assignment is greater than the maximum 

likelihood dose, assigning an unmonitored external dose of 100 mrem/yr as only photon dose is still 

considered to be favorable to claimants for the leukemia cases based on the results of this evaluation.

ATTACHMENT B

BASIS FOR THE UNMONITORED EXTERNAL DOSE ASSIGNMENT

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Document No. ORAUT-TKBS-0029-6 Revision No. 01 Effective Date: 04/28/2011 Page 59 of 59

Table B-2. Evaluation of POCs associated with assigning unmonitored 

doses as either photon or neutron dose.

Applicable cancer models

Worst-case 

external dose 

organ selection

Photon

dose POC

(%)

Neutron

dose POC

(%)

Acute lymphocytic leukemia Red bone marrow 17.89 11.01

Acute myeloid leukemia Red bone marrow 10.59 11.83

All digestive Stomach 4.34 2.56

All male genitalia Testes 3.57 2.21

Bladder Bladder 4.68 3.52

Bone Bone surfaces 4.98 3.38

Breast Breast (female) 7.58 4.04

Chronic myeloid leukemia Red bone marrow 12.63 12.71

Colon Colon 5.33 3.96

Connective tissue Thyroid 5.36 3.45

Esophagus Esophagus 4.95 3.99

Eye Eye 4.72 3.08

Female genitalia (less ovary) Bladder 0.04 0.03

Gallbladder Bladder 11.66 6.65

Leukemia (less CLL) Red bone marrow 15.69 18.34

Liver Liver 17.14 10.61

Lung Thyroid 5.72 3.03

Lymphoma and multiple myeloma Thyroid 3.43 1.77

Malignant melanoma Skin 10.09 7.63

Nervous system Thyroid 2.41 1.45

Non-melanoma (BCC) Skin 10.20 7.82

Non-melanoma (SCC) Thyroid 0.33 0.15

Oral cavity and pharynx Thyroid 2.59 1.54

Other and ill-defined sites Thyroid 5.19 3.13

Other endocrine Thyroid 6.30 4.00

Other respiratory Thyroid 2.26 1.05

Ovary Bladder 5.42 3.84

Pancreas Stomach 2.47 1.60

Rectum Colon 1.70 1.17

Stomach Stomach 7.95 4.81

Thyroid Thyroid 16.08 7.10

Urinary organs (less bladder) Testes 6.24 4.24

ATTACHMENT B

BASIS FOR THE UNMONITORED EXTERNAL DOSE ASSIGNMENT

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